ML092920190
| ML092920190 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 10/17/2009 |
| From: | NRC/RGN-II |
| To: | Virginia Electric & Power Co (VEPCO) |
| References | |
| 50-280/09-301, 50-281/09-301 | |
| Download: ML092920190 (158) | |
Text
1. 000008AA2.20 2 Current conditions on Unit 1 are as follows:
Reactor power is 100%.
All plant systems and components are in a normal configuration.
RCS hot and cold leg temperatures are stable.
Containment pressure is 12.2 psia and rising slowly.
Pressurizer pressure is 2005 psia and slowly lowering.
Pressurizer level is 59% and rising.
Charging flow is lowering in automatic.
Which ONE of the following containment leak locations correlates to the above indications?
A. RCS cold leg B. Main steam line C. Reactor vessel head D. Pressurizer vapor space K/A Pressurizer Vapor Space Accident: Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: The effect of an open PORV or code safety, based on observation of plant parameters.
K/A Match Analysis The applicant must understand that the indications in the stem are indicative of a vapor space accident, thereby testing the required knowledge of interpreting plant parameters to understand that a vapor space LOCA is occurring.
Answer Choice Analysis A. Incorrect, but plausible if the applicant identifies an increase in containment pressure with LOCA condition but fails to identify that RCS temperatures are NOT decreasing as expected.
B. Incorrect, but plausible if the applicant identifies MSL as being inside containment and incorrectly believes that the rapid cooldown and subsequent depressurization
will result in a lower pressure than a loss of coolant from the RCS.
C. Incorrect, but plausible if the applicant identifies the Rx Vessel as a steam space because both spray valves are open causing a reactor trip on OTdelta-T. or Low Pressurizer Pressure, followed by an SI.
D. Correct.
Supporting References ND-93.3-LP-5, Pressurizer Pressure Control, Rev. 9 References Provided to Applicant none Typically we have two orifices in service. Recommend modification of stem.
Typo in distractor 'C' plausibility statement.
Answer: D
- 2. 000009EK2.03 3 Unit 1 Initial Conditions:
A small break loss-of-coolant accident (SBLOCA) occurred from 100%
power.
Operators are performing steps in 1-ES-1.2, "POST LOCA COOLDOWN AND DEPRESSURIZATION."
A controlled RCS cooldown has been initiated at approximately 90 °F/hr.
All Steam Generator (S/G) narrow range (NR) levels are approximately 45% and STABLE.
All S/G pressures are approximately 650 psig and DECREASING.
Auxiliary Feedwater Flow to EACH steam generator is initially 200 gpm.
Current conditions:
Operators stopped the CHG pump flowing to the alternate header.
Operators then "paused" for approximately five (5) minutes after stopping the CHG pump to allow RCS pressure to stabilize or increase before taking further actions to reduce SI flow.
Following safety injection flow reduction, cooldown rate was calculated to be 77°F/hr.
No operator actions were performed during the "pause."
Based on the conditions at the end of the "pause" (approximately five (5) minutes after stopping the CHG pump), which ONE of the following predicts:
(1) S/G NR level response AND (2) S/G pressure response?
A. (1) levels will be INCREASING.
(2) pressures will be DECREASING.
B. (1) levels will be INCREASING.
(2) pressures will be STABLE at a lower value.
C. (1) levels will be STABLE at the same value.
(2) pressures will be DECREASING.
D. (1) levels will be STABLE at the same value.
(2) pressures will be STABLE at a lower value.
K/A Small Break LOCA Knowledge of the interrelations between the small break LOCA and the following:
S/Gs. (CFR: 41.7/45.7) (RO - 3.0)
K/A Match Analysis The question forces the applicant to identify an expected set of S/G parameters during a SBLOCA accident condition.
Answer Choice Analysis A. CORRECT. (1) With no adjustment to AFW flow control valves, AFW flow will INCREASE as the S/Gs continue to depressurize. This will tend to cause S/G levels to INCREASE. (2) Even with no adjustment to steam dump controls (or atmospheric controls), the S/Gs will continue to depressurize (perhaps at a slower rate)
B. INCORRECT. (1) is Correct, (2) is INCORRECT. Even with no adjustment to steam dump controls, the S/Gs will continue to depressurize. The distractor is plausible if the candidate loses sight that the cooldown will continue during the "pause."
C. INCORRECT. (1) is incorrect, levels will rise as AFW increases against a lowering S/G backpressure. Distractor is plausible if the candidate focuses on the fact that levels were initially stable, and assumes that they will stay that way.
(2) is the correct option.
D. INCORRECT. Both (1) and (2) are incorrect; see above analyses.
Supporting References
-Surry procedure 1-ES-1.2, "POST LOCA COOLDOWN AND DEPRESSURIZATION," especially the continuous action page, and p.10 (steps 16 and 17: SI reduction steps)
-Steam Tables (for parameter validity).
References Provided to Applicant Steam Tables.
Modify to include initial AFW flow rate?
Need to focus candidates on AFW flow and that the cooldown continues.
Answer: A
- 3. 000011EK3.12 5 Unit 1 initial plant conditions:
Time = 0800 Reactor power = 100%
Current plant conditions:
Time = 0845 RCS pressure = 700 psig decreasing RCS Subcooling = 20°F decreasing Safety Injection Flow = 225 gpm to each loop Based on the above conditions after transition to 1-E-1 (Loss of Reactor or Secondary Coolant): (1) which ONE of the following actions are directed by 1-E-1 with regard to RCPs and (2) what is the reason for that action?
A. (1) Secure RCPs (2) To reduce the depletion of RCS water inventory.
B. (1) Secure RCPs (2) To prevent the possibility of flywheel fracture if the pump continues to operate without coolant.
C. (1) Maintain RCPs operating (2) They provide core cooling by pumping a 2 phase mixture through the core and loops.
D. (1) Maintain RCPs operating
(2) To prevent phase separation in the core region which could lead to core uncovery.
K/A Large Break LOCA / 3 Knowledge of the reasons for the following responses as they apply to the Large Break LOCA: Actions contained EOP for emergency LOCA (large break).
K/A Match Analysis Question requires knowledge of RCP trip criteria during a LB LOCA.
Answer Choice Analysis A. Correct. Per 1-E-1, Step 1, if RCS subcooling is less that 30 0F, Stop all RCPs. As stated in Lesson Plan ND-95.3-LP-7, E.1.b, for large break LOCAs, the operation of the RCPs has little, if any effect during mitigation and recovery.
B. Incorrect. 1st part is correct. second part is plausible because in the lesson plan (same paragraph as in A), the RCP is maintained running for a short period of time to prevent flywheel fracture. However it would seem more likely to fracture during two phase operation.
C. Incorrect. With subcooling < 30 0F, RCPs are secured. Plausible because if subcooling were >30 0F, it would be correct.
D. Incorrect, With subcooling < 30 0F, RCPs are secured. Plausible because if subcooling were >30 0F, it could be correct depending on additional criteria.
Supporting References ND-95.3 LP-7, Obj A, E References Provided to Applicant none Need to adjust time for realism and progression through EOPs.
SI flow is required to meet RCP trip criteria that is tested.
Answer: A
- 4. 000015AA2.09 2 Current Unit 1 plant conditions:
- Reactor power is at 33%
The latest temperature readings obtained from TR-1-448 are as follows:
Temperatures (°F) RCP 'A' RCP 'B' RCP 'C' Upper thrust bearing 181 178 163 Lower thrust bearing 173 183 172 Upper radial bearing 143 163 146 Lower radial bearing 172 189 158 Motor stator 285 273 302 Lower bearing seal water 153 183 167 Seal water 195 184 185 Given the above temperature readings, which one of the following correctly states the RCP, if any, that exceeds an ACTION LEVEL limit in Attachment 2, RCP Parameters, of 1-AP-9.00, RCP Abnormal Conditions?
A. RCP A.
B. RCP B.
C. RCP C.
D. No RCPs are exceeding an ACTION LEVEL limit.
K/A Reactor Coolant Pump Malfunctions Ability to determine and interpret the following as it applies to Reactor Coolant Pump Malfunctions (Loss of RC Flow): When to secure RCPs on high stator temperatures.
K/A Match Analysis Requires applicant to recognize that the RCP C motor stator temperature has exceeded the temperature that requires tripping of the pump.
Answer Choice Analysis A. In-Correct but plausible since the alarm point for high seal water temperature is 195ºF. However, the temperature where the ACTION LEVEL is reached occurs at 225ºF.
B. In-Correct but plausible since the alarm point for lower thrust bearing temperature is 175ºF. However, the temperature where the ACTION LEVEL is reached occurs
at 195ºF.
C. Correct - The ACTION LEVEL for motor stator is when the temperature reaches 300ºF.
D. In-Correct but plausible if the applicant was not familiar with the ACTION LEVEL temperatures.
Supporting References ND-88.1-LP-6, Reactor Coolant Pumps, Rev. 019 - Obj E 1-AP-9.00, RCP Abnormal Conditions, Rev. 20 References Provided to Applicant none Answer: C
- 5. 000022AG2.4.11 2 Initial plant conditions are as follows:
Unit 1 is at 100% power.
Unit 2 is at 100% power.
1-CH-P-1B, Unit 1 B Charging pump, is out of service with motor removed.
1-CH-P-1C, Unit 1 C Charging pump, is running on its alternate power supply.
Current plant conditions are as follows:
1-CH-P-1C has tripped.
Attempts to start 1-CH-P-1A have been unsuccessful.
The crew has entered 1-AP-8.00 Loss of Normal Charging Flow.
No indications of Unit 1 charging system leakage are observed.
All Unit 2 Charging pumps are operable.
RCP seal injection flow is zero.
Component cooling water flow to the thermal barrier is normal.
Given the above conditions, which ONE of the following would be consistent with the actions required by 1-AP-8.00?
A. Trip Unit 1 reactor. Do NOT trip Unit 2 reactor.
Cross-connect charging with Unit 2 per 1-AP-8.00.
B. Trip Unit 1 reactor. Do NOT trip Unit 2 reactor.
Cross-connecting charging with Unit 2 is NOT permitted per 1-AP-8.00.
C. Trip Unit 1 and Unit 2 reactors.
Cross-connect charging with Unit 2 per 1-AP-8.00.
D. Trip Unit 1 and Unit 2 reactors.
Cross-connecting charging with Unit 2 is NOT permitted per 1-AP-8.00.
K/A:
022AG.2.4.11 Loss of Reactor Coolant Makeup Knowledge of abnormal condition procedures as it relates to: Loss of Reactor Coolant Makeup.
K/A MATCH ANALYSIS:
The question focuses on recognizing the conditions within AP-8.00 for cross-connecting with the other units Charging System and the associated actions.
ANSWER CHOICE ANALYSIS:
A. In-Correct but plausible since these are required actions for cross-connecting with the other units charging system. However, Unit 2 is also required to be tripped.
B. In-Correct but plausible since tripping only the affected unit would be correct if Unit 1s charging system had a leak. Also, all the required conditions are in place to allow cross-connecting charging system.
C. Correct - Based on the loss of all charging on Unit 1, no signs of leaks, and an intact system as well as the availability of charging pumps on Unit 2, this choice contains all the actions per AP-8.00 for cross-connecting the charging systems including tripping both units.
D. In-Correct but plausible since tripping both units would be correct. Also, all the required conditions are in place to allow cross-connecting charging system.
REFERENCES:
ND-88.3-LP-2, Charging and Letdown, Rev. 015 AP-8.00, Loss of Normal Charging Flow, Rev. 011 Modifications made to correct mark numbers and charging pumps out of service.
Answer: C
- 6. 000025AK1.01 3 Unit conditions are as follows:
- A unit shutdown is in progress due to increased RCS leakage
- RCS Temperature is 300°F
- RCS Pressure is 300 psig
- The RCS is solid RH-P-1A ("A" RHR Pump) is in service with 1-RH-E-1A ("A" RHR Heat Exchanger)
Which ONE of the following would INITIALLY occur if 1-RH-P-1A were to trip on overcurrent?
A. RCS Pressure would increase B. 1-CH-PCV-1145 (Letdown Pressure Control Valve) would open C. Charging flow would increase D. CC head tank level would increase K/A Loss of RHR System.
Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation.
K/A Match Analysis Requires knowledge of operational impact resulting from a loss of RHR control during solid plant condition.
Answer Choice Analysis:
- 1. CORRECT.
B. INCORRECT. Plausible if candidate fails to understand operation of PCV-1145 during solid plant operations.
C. INCORRECT. Plausible if candidate fails to understand that charging is maintained in MANUAL control during solid plant operations.
D. INCORRECT. Plausible if candidate cannot determine cause/effect relationship from a loss of RHR on the CC system.
Supporting
References:
ND-88.2-LP-1 AP-27.00 References Provided to Applicant:
None.
Answer: A
- 7. 000027AK3.03 2 Unit 1 initial conditions:
Time = 1000 Reactor power = 100%
1-RC-PCV-1455B (Pzr Spray Valve ) fails open RCS pressure = 2100 psig decreasing Current conditions:
Time = 1001 RCS pressure = 1900 psig decreasing 1-E-0 REACTOR TRIP OR SAFETY INJECTION in progress Based on the above conditions: (1) state which ONE of the following actions is directed by step 7 of 1-E-0 if the Pzr Spray Valve can not be closed and (2) state the reason why?
A. (1) Secure RCP A (2) To stop the RCS depressurization.
B. (1) Secure RCP A (2) To prevent inadvertent Safety Injection.
C. (1) Secure RCP C (2) To stop the RCS depressurization.
D. (1) Secure RCP C (2) To prevent inadvertent Safety Injection.
K/A Pressurizer Pressure Control System Malfunction / 3 Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: Actions contained in the EOP for Pzr PCS malfunction.
K/A Match Analysis Requires knowledge of reasons for actions taken in the EOP for Pzr Pressure control malfunction.
Answer Choice Analysis A. Incorrect: 1RCPCV 1455B is supplied by the 1C RCP. 2nd part is correct.
B. Incorrect: 1RCPCV 1455B is supplied by the 1C RCP. 2nd part is incorrect
because it will not prevent SI. 2nd part is plausible because it will delay the SI initiation.
C. Correct: 1RCPCV 1455B is supplied by the 1C RCP. The RCP is secured to reduce the spray flow and therefore the rate of RCS pressure decrease.
D. Incorrect: 1st part is correct. 2nd part is incorrect because it will not prevent SI.
2nd part is plausible because it will delay the SIinitiation.
Supporting References 1-E-0 Step 7 LP ND-95.3-LP3, Obj A References Provided to Applicant none Check on simulator and adjust distractors as needed. Otherwise OK Answer: C
- 8. 000029EA1.02 2 Unit 1 Initial Conditions:
An Anticipated Transient Without SCRAM (ATWS) occurred at 100%
power.
The reactor remains at power. An operator is inserting control rods in manual.
Safety Injection is NOT actuated.
Charging flow was verified to be 77 GPM, and the BATP was placed in FAST.
Current conditions:
1-CH-MOV-1350, Emergency Borate MOV, will not open. An operator reports that the valve appears to be mechanically bound.
Neither Pressurizer PORV automatically opened when RCS pressure rose above 2335 psig. An operator was able to manually open ONLY one Pressurizer PORV to control RCS pressure.
Based on the current conditions, which ONE of the following correctly identifies (1) the next required action to initiate emergency boration, in accordance with 1-FR-S.1, "RESPONSE TO NUCLEAR POWER GENERATION/ATWS," AND (2) the required action to operate the PORV as specified in FR-S.1?
A. (1) Manually actuate SI to provide for maximum flowrate injection into the RCS.
(2) Allow the RCS pressure to lower to 2210 psig before closing the PORV.
B. (1) Place switches for CH-MOV-1115B and -1115D to OPEN and switches for CH-MOV-1115C and -1115E to CLOSE.
(2) Allow the RCS pressure to lower to 2210 psig before closing the PORV.
C. (1) Manually actuate SI to provide for maximum flowrate injection into the RCS.
(2) Close the PORV when RCS pressure equals 2335 psig and lowering.
D. (1) Place switches for CH-MOV-1115B and -1115D to OPEN and switches for CH-MOV-1115C and -1115E to CLOSE.
(2) Close the PORV when RCS pressure equals 2335 psig and lowering.
K/A Anticipated Transient Without SCRAM (ATWS)
Ability to operate and monitor the following as they apply to an ATWS: charging pump suction valves from RWST operating switch.
(CFR 41.7/45.5/45.6) (RO - 3.6)
K/A Match Analysis The question requires the RO applicant to understand the intent of FR-S.1 step 5, and to know how the valve switches identified in the answers correspond to aligning CHG pump suction to the RWST.
Answer Choice Analysis A. INCORRECT. (1) The distractor is plausible because actuating SI may trip the reactor, and will provide for increased flowrate into the RCS; however the Surry lesson plan for FR-S.1 specifically states that the operations team should NOT manually actuate SI in an attempt to get the reactor to trip. The SI would also trip off the MBFPs, which may be providing the heat sink function (not stated in the question stem). (2) This distractor is also incorrect, because the step as read in FR-S.1 states: " ...open PRZR PORVs and block valves as necessary until PRZR pressure less than 2335 psig." This part is plausible, because the WOG background document to FR-S.1 specifically states that the intent of this step is to close the PORV when primary pressure drops to 200 psi below the PORV pressure setpoint to allow increased injection flow. Surry just happens to not follow this step as written in the WOG.
B. INCORRECT. Part (1) is the correct answer; FR-S.1 step 5.b RNO states that the operator is required to either "Manually align CHG pump suctions to RWST OR put the blender mode selector switch in the BORATE position and start the blender." The step is re-written to specify the individual switch
combinations to fully meet the K/A. Part (2) is incorrect, as detailed in the analysis of A. above.
C. INCORRECT. See analyses for A. and B. above.
D. CORRECT. See analyses for A. and B. above.
Supporting References
-Surry procedure 1-FR-S.1, "Response to Nuclear Power Generation," rev 25 p.
3; step 5.b) RNO.
-Surry lesson plan ND-95.3-LP-36, rev. 13, especially p. 13
-WOG background document for FR-S.1, HP-Rev 2, p. 80 and 81.
References Provided to Applicant none Question almost requires memorization of step 5 of FR-S.1. But the first set of distractors test the intent of the step vice step memorization. The second set of distractors does require memorization...
Answer: D
- 9. 000038EA1.21 2 Unit 1 initial plant conditions:
Reactor power = 100%
Pzr level = 52 % and stable VCT Level = 40% and decreasing slowly 1-AP-16.00 (Excessive RCS Leakage) initiated Current Unit 1 conditions:
Leak determined to be Steam Generator Tube Leak in the 1A SG = 39 gpm and increasing slowly The team has initiated 1-AP-24.00 (Minor Steam Generator Tube Leak)
Based on the above conditions: (1) how will charging pump amps change as 1-CH-FCV-1122 (Charging Flow Control Valve) opens to maintain pressurizer level and (2) is a reactor trip required at this instant in time per 1-AP-16.00 or 1-AP-24.00?
A. (1) pump amps will increase (2) Yes B. (1) pump amps will increase (2) No
C. (1) pump amps will decrease (2) Yes D. (1) pump amps will decrease (2) No K/A Steam Gen. Tube Rupture / 3 Ability to operate and monitor the following as they apply to a SGTR. Charging pump ammeter and running indicator.
K/A Match Analysis Requires knowledge of how charging is controlled during a SGTR and charging pump indications.
Answer Choice Analysis A. Incorrect: 1st part is correct. As charging flow increases to maintain pressurizer level, more work is performed so pump amps increase. 2nd part is plausible because the leak rate is significant enough to require a rapid plant shutdown.
B. Correct. As charging flow increases to maintain pressurizer level, more work is performed so pump amps increase. The leak rate that requires a reactor trip per AP16/24 is 50 gpm.
C. Incorrect: 1st part is plausible because pump discharge pressure will decrease as the discharge valve opens (common misconception) 2nd part is plausible because the leak rate is significant enough to require a rapid plant shutdown.
D. Incorrect: 1st part is plausible because pump discharge pressure will decrease as the discharge valve opens (common misconception) 2nd part is correct.
Supporting References AP/16 Step 4, AP/24 Step 1, ND-93.3-LP-7 D.2.a, Obj B References Provided to Applicant none
May need to lower flow of leak, as it is close to the trip value and depending on thought process, they may opt to trip. Could the leak rate be lowered to 19 gpm?
Or ask "at this instant in time...is a trip required.?"
Answer: B
- 10. 000040AA2.05 1 Which ONE of the following completes the below statements?
(1) The parameters used for SI Termination criteria in 1-E-2, "FAULTED STEAM GENERATOR ISOLATION," are RCS subcooling AND ___________________
AND (2) The parameters used for SI Termination criteria in 1-ECA-2.1, "UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS?" are RCS subcooling AND
________________?
A. (1) E-2: ONLY RCS pressure and PRZR level.
(2) ECA-2.1: RVLIS indication greater than values specified in a table (based upon number of running RCPs).
B. (1) E-2: ONLY RCS pressure and PRZR level.
(2) ECA-2.1: RCS pressure and PRZR level.
C. (1) E-2: Secondary heat sink, RCS pressure, and PRZR level.
(2) ECA-2.1: RVLIS indication greater than values specified in a table (based upon number of running RCPs).
D. (1) E-2: Secondary heat sink, RCS pressure, and PRZR level.
(2) ECA-2.1: RCS pressure, and PRZR level.
K/A Steam Line Rupture Ability to determine and interpret the following as they apply to the Steam Line Rupture: When ESFAS systems may be secured.
(CFR: 43.5/45.13) (RO - 4.1)
K/A Match Analysis The question gives the applicant an opportunity to demonstrate understanding of the differences behind the various criteria used for SI Termination (when ESFAS systems may be secured) given various cases of steam line rupture (the
'standard' E-2 case and the 'all faulted generators' case of ECA-2.1).
Answer Choice Analysis A. INCORRECT. (1) Lists the SI Termination criteria in E-2 but omits the secondary heat sink portion. (2) is the SI Termination criteria as used in the procedure FR-P.1, "RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION." This is a plausible, yet incorrect, answer, because the large RCS cooldown caused by an 'all faulted generators' condition (as in ECA-2.1) may cause a RED or ORANGE path in the INTEGRITY critical safety function. However, this is not what the question stem asks for.
B. INCORRECT. (1) incorrect as per the analysis of A. above, (2) is the correct SI Termination criteria from ECA-2.1; basically the exact same as E-2, without the heat sink portion.
C. INCORRECT. (1) is correct, (2) is the FR-P.1 criteria.
D. CORRECT. see above analysis.
Supporting References
-Surry procedure 1-E-2, "FAULTED STEAM GENERATOR ISOLATION," rev 15,
- p. 6 (step 8).
-Surry procedure 1-ECA-2.1, "UNCONTROLLED DEPRESSURIZATION OF ALL STEAM GENERATORS," rev. 29, p. 9 (step 12), p. 12-15 (step 17-23)
-Surry procedure 1-FR-P.1, "RESPONSE TO IMMINENT PRESSURIZED THERMAL SHOCK CONDITION," rev 16, p. 5 (step 6)
References Provided to Applicant None Challenging.
Answer: D
- 11. 000055EK1.02 1 Unit 1 initial conditions:
Station Blackout occurs 1-ECA-0.0 (LOSS OF ALL AC POWER) in progress SGs are to be depressurized to allow SI accumulators to inject into the RCS.
Based on the above conditions, which ONE of the following: (1) correctly states the maximum cooldown rate allowed during this depressurization and (2) the basis for that
rate ?
A. (1) 25 oF/Hr (2) To prevent a steam bubble from forming in the reactor vessel head.
B. (1) 25 0F/Hr (2) To minimize RCS inventory loss.
C. (1) 100 0F/Hr (2) To prevent a steam bubble from forming in the reactor vessel head.
D. (1) 100 0F/Hr (2) To minimize RCS inventory loss.
K/A Station Blackout / 6 Knowledge of the operation implications of the following concepts as they apply to the Station Blackout: Natural circulation cooling.
K/A Match Analysis Requires knowledge of limitations placed on the plant during natural circulation as a result of a station blackout. RO knowledge because information is contained in EOP Note & Cautions.
Answer Choice Analysis A. Incorrect: 1st part is plausible because 25 0F/Hr in the cooldown rate limit for natural circulation cooldown. 2nd part is plausible because it is the reason for limiting the cooldown rate during a natural circulation cooldown.
B. Incorrect: 1st part is plausible because 25 0F/Hr in the cooldown rate limit for natural circulation cooldown. 2nd part is correct.
C. Incorrect: 1st part is correct. 2nd part is plausible because it is the reason for limiting the cooldown rate during a natural circulation cooldown.
D. Correct: Note before ECA 0-0 Step 21 states that the SGs should be depressurized at the maximum controllable rate, not to exceed an RCS cooldown rate of 100 0F/Hr. This to minimize RCS Inventory loss.
Supporting References ND-95.3-LP-17 Step 33. Obj: A 1-ECA-0.0 Step 21 Caution 1-ES-0.2 (Nat Circ Cooldown) Step 6 References Provided to Applicant
none Answer: D
- 12. 000056AA1.29 3 Initial Conditions:
Surry Unit 2 is in day 25 of a scheduled refueling outage.
A Loss of All AC Power occurred on Unit 1, which was operating at 100%
power.
Control room operators implemented 1-ECA-0.0, "LOSS OF ALL AC POWER."
Safety Injection (SI) initiated on Unit 1.
Power was restored to one Unit 1 safeguards power train from an Emergency Diesel Generator.
Current conditions:
Unit 1 control room operators are implementing 1-ECA-0.2, "LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED," and are performing the step that directs restoration of SW to CC HXs IAW 0-AP-12.01, "LOSS OF INTAKE CANAL LEVEL."
Service water is isolated to all recirculation spray (RS) heat exchangers on both units.
2 ESW pumps are running.
Note: In 0-AP-12.01, "time zero" is defined as that time intake canal level reaches 23.5 FT.
Based on the current conditions, which ONE of the following correctly identifies (1) the 0-AP-12.01 restriction, if any, on CC HX SW flow 15 minutes after "time zero" AND (2) the 0-AP-12.01 restriction on CC HX SW flow after nine hours have elapsed from "time zero?"
(Reference provided)
A. (1) There are no restrictions on CC HX SW flow.
(2) Crosstie CC. 3 CC HXs allowed with SW outlet valves 19 turns open for each HX.
B. (1) There are no restrictions on CC HX SW flow.
(2) 2 CC HXs allowed with SW outlet valves 19 turns open for each HX.
C. (1) Maximum allowable flow is one CC HX outlet SW valve fully open.
(2) Crosstie CC. 3 CC HXs allowed with SW outlet valves 19 turns open for each HX.
D. (1) Maximum allowable flow is one CC HX outlet SW valve fully open.
(2) 2 CC HXs allowed with SW outlet valves 19 turns open for each HX.
K/A Loss of Offsite Power Ability to operate and/or monitor the following as they apply to the Loss of Offsite Power: CCW heat exchanger temperature control valves.
(CFR: 41.7/45.5/45.6) (RO - 2.7)
K/A Match Analysis The question requires the RO applicant to demonstrate knowledge to operate CCW heat exchanger SW valves in a post-loss of offsite power situation.
Answer Choice Analysis A. Correct.
B. Incorrect.
C. Incorrect.
D. Incorrect.
Supporting References ECA-0.2, "LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED,"
rev. 15, p. 9. AP-12.01, "LOSS OF INTAKE CANAL LEVEL," rev. 25, especially p. 6-9.
References Provided to Applicant
- AP-12.01 pages 6, 7, and 8.
Need to pick better times. At one hour and At 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are transition times, add 15 minutes on to each item.
Need pages 9 and 10 too.
Answer: A
- 13. 000057AA2.20 2 Unit 2 is performing a plant shutdown with the reactor at 5% power when a loss of Vital Bus 1 occurs.
Which one of the following correctly describes:
(1) the direct effect of the loss of Vital Bus I on the RPS system and (2) if a trip or shutdown were to occur, the effect on the re-instatement of the Source Range NIs?
A. (1) An RPS trip signal is generated (2) Re-instatement of SRNIs will occur automatically.
B. (1) An RPS trip signal is generated (2) Re-instatement of SRNIs will NOT occur automatically.
C. (1) An RPS trip signal is NOT generated (2) Re-instatement of SRNIs will occur automatically.
D. (1) An RPS trip signal is NOT generated (2) Re-instatement of SRNIs will NOT occur automatically.
K/A Loss of Vital AC Inst. Bus.
Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: Interlocks in effect on loss of ac vital electrical instrument bus that must be bypassed to restore normal equipment operation.
K/A Match Analysis Requires knowledge of how a loss of vital power affects interlocks.
Answer Choice Analysis A. Incorrect: 1st part is correct. 2nd part is plausible because for Vital Bus 3 or 4, it would be correct.
B. Correct. The IR channel signal will fail HIGH on a loss of Vital bus I. With power <
P 10 this generates a high power trip (1/2) > 35% IR channels. The IR channel failed high will also prevent the 2/2 signal <5 x 10-11 amps required to automatically energize the SR NIs when shutting down the reactor.
C. Incorrect: 1st part is plausible because if power were greater than 10%, it would be correct. 2nd part is plausible because for Vital Bus 3 or 4, it would be correct.
D. Incorrect: 1st part is plausible because if power were greater than 10%, it would be
correct. 2nd part is correct.
Supporting References ND.93.2-LP-3 (Intermediate Range NIs) Obj: D ND-93.2-LP-2 (Source Range NIs) Obj: C ND-93.3-LP-5 (Vital Power)
References Provided to Applicant none Verify with the new words.
Answer: B
- 14. 000058AG2.1.23 3 Unit 1 Initial Conditions:
100% power.
Current conditions:
A complete loss of DC bus 1B has occurred.
Based on the current conditions, which ONE of the following correctly identifies (1) the required action in 1-AP-10.06, "LOSS OF DC POWER," for the generator output breakers, AND (2) the subsequent impact on Reactor Coolant Pump (RCP) operations following completion of the immediate actions of 1-E-0 (Reactor Trip or Safety Injection)?
A. (1) A control room operator is NOT required to manually open the Generator output breakers.
(2) 'A' RCP will stop, 'B' and 'C' RCPs will remain running.
B. (1) A control room operator is NOT required to manually open the Generator output breakers.
(2) 'A' RCP will remain running, 'B' and 'C' RCPs will stop.
C. (1) A control room operator is required to manually open the Generator output breakers.
(2) 'A' RCP will stop, 'B' and 'C' RCPs will remain running.
D. (1) A control room operator is required to manually open the Generator output breakers.
(2) 'A' RCP will remain running, 'B' and 'C' RCPs will stop.
K/A Loss of DC Power Ability to perform specific system and integrated plant procedures during all modes of plant operation (as related to Loss of DC Power)
(CFR: 41.10/43.5/45.2/45.6) (RO - 4.3)
K/A Match Analysis The RO applicant must demonstrate knowledge of the effects of a loss of 1B DC bus, and how recovery actions are complicated by output generator and RCP operation.
Answer Choice Analysis NOTE TO SURRY: Please validate this question against the action verb "VERIFY" in step 3 of 1-AP-10.06--e.g. if operators automatically match flags and always perform the RNO step, this question may need to be modified.
A. INCORRECT. Incorrect about the generator output breakers, but plausible because it would be true for a loss of 1A DC Bus. Incorrect about the RCP operation, but plausible because it would be true for a loss of 1A DC Bus.
B. INCORRECT. See analysis for A. above. Correct RCP operations, as specified in 125 VDC lesson plan on p. 18.
C. INCORRECT. Correct output breaker description, as specified in the lesson plan on p. 17. Incorrect RCP ops.
D. CORRECT. See above analyses.
Supporting References
- Surry lesson plan ND-90.3-LP-6, "125VDC DISTRIBUTION," rev. 14, p. 12-20. AP-10.06, "LOSS OF DC POWER," rev. 13, p.
References Provided to Applicant
none
'B' and 'C' will remain running, until a the MG voltage degrades (the breakers cannot open). The answer for this question depends on timing of answering the question, adding the word subsequently should eliminate the timing aspect.
Answer: D
- 15. 00005AA2.01 2 Unit 1 initial conditions; Reactor power = 90% and decreasing Shutdown in progress for refueling Current plant conditions:
Reactor power = 70% and decreasing 1G-B5 COMPUTER PRINTOUT ROD CONT SYS in alarm QPTR = 1.05% and increasing The power decrease is stopped (1) Based on the above indications, which ONE of the following conditions has occurred and (2) what is the minimum QPTR that requires operator action when exceeded IAW Tech Spec 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS?
A. (1) Stuck control rod (2) 2%
B. (1) Stuck control rod (2) 5%
C. (1) Control bank inserted past insertion limit (2) 2%
D. (1) Control bank inserted past insertion limit (2) 5%
K/A Inoperable/Stuck Control Rod Ability to determine and interpret the following as they apply to the Inoperable/Stuck Control Rod: Stuck or inoperable rod from in-core and ex-core NIs, in-core or loop temperature measurements.
K/A Match Analysis Requires applicant to interpret ex-core NI data (QPTR) to determine that a control rod is stuck.
Answer Choice Analysis A. Correct: A stuck control will cause QPTR to increase as the rest of the control rod bank moves. Per Tech Spec 3.12, QPTR of > 2% requires operator action.
B. Incorrect: 1st part is correct. 2nd part is incorrect because per Tech Spec 3.12, a QPTR of > 2% requires operator action. 2nd part is plausible because per T.S. 3.12 Bases, a measured QPTR of 2% ensures an actual QPTR due to instrument sensitivity and error could be 5%.
C. Incorrect: 1st part is incorrect because a control rod bank moving together will not produce QPTR issues. 1st part is plausible because a control rod bank inserted past its insertion limit will cause neutron flux distribution issues (flux difference).
2nd part is correct.
D. Incorrect: 1st part is incorrect because a control rod bank moving together will not produce QPTR issues. 1st part is plausible because a control rod bank inserted past its insertion limit will cause neutron flux distribution issues (flux difference).
2nd part is incorrect because per Tech Spec 3.12, a QPTR of > 2% requires operator action. 2nd part is plausible because per T.S. 3.12 Bases, a measured QPTR of 2% ensures an actual QPTR due to instrument sensitivity and error could be 5%.
Supporting References 1G-B5 COMPUTER PRINTOUT ROD CONT SYS TS 3.12 References Provided to Applicant none Minor admin changes Answer: A
- 16. 000065AA1.03 3 Following a complete loss of instrument air and subsequent restoration, which ONE of the following components will require manual re-alignment to support the start of 1-RC-P-1C?
A. 1-CC-TV-105C, RCP C Cooler CC Return Trip Valve B. 1-CC-TV-120C, RCP C Thermal Barrier Component Cooling Water System Trip Valve C. 1-RC-HCV-1303C, RCP C Seal Leakoff Isolation Valve D. 1-CH-MOV-1381, RCP Seal Return Valve K/A Loss of Instrument Air Ability to operate and / or monitor the following as they apply to the Loss of Instrument Air: Restoration of sytems served by instrument air when pressure is regained.
K/A Match Analysis Requires knowledge of how RCP support equipment is impacted upon a loss of IA and how restoration is accomplished when IA is returned.
Answer Choice Analysis A. Correct.
B. Incorrect: Plausible as 1-CC-TV-120C fails open. This is the only trip valve that fails open on a loss of instrument air.
C. Incorrect: Plausible as the candidate may believe that the valve fails closed to maintain RCS inventory (similar to letdown, RHR letdown, loop drains, etc...).
D. Incorrect: Plausible as the candidate may believe that since this is a Phase I isolation valve that it will fail in the position required by safety injection. A significant amount of SI re-alignment is accomplished by venting air off pressure switches, but not for this valve.
Supporting References 1-OP-RC-001(RCP Operation)
References Provided to Applicant none Answer: A
- 17. 000077AA2.07 1 Unit 1 Initial Conditions:
100% Power Both Megawatts and Megavars start to oscillate.
System Operator reports that the regional grid is experiencing dynamic instabilities.
Current conditions:
230 KV system voltage is 212 KV 500 KV system voltage is 492 KV Based on the current conditions, which one of the following identifies (1) the required operator action for continued voltage regulator operation, as specified in 0-AP-10.18, "RESPONSE TO GRID INSTABILITY," AND (2) the required entry into Technical Specification (TS) 3.16, "Emergency Power System?"
A. (1) Place the voltage regulator - in MANUAL.
(2) Entry into TS 3.16 is required ONLY for the 230 KV system. Entry into TS 3.16 for the 500 KV system is NOT required.
B. (1) Place the voltage regulator - in MANUAL.
(2) Entry into TS 3.16 is required for BOTH the 230 KV system AND the 500 KV system.
C. (1) Verify the voltage regulator - in AUTO, or place the voltage regulator in AUTO if possible.
(2) Entry into TS 3.16 is required for BOTH the 230 KV system AND the 500 KV system.
D. (1) Verify the voltage regulator - in AUTO, or place the voltage regulator in AUTO if possible.
(2) Entry into TS 3.16 is required ONLY for the 230 KV system. Entry into TS 3.16 for the 500 KV system is NOT required.
K/A Generator Voltage and Electric Grid Disturbances Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: operational status of offsite circuit.
(CFR: 41.5 and 43.5/45.5,45.7, and 45.8) (RO - 3.2)
K/A Match Analysis
The question requires the candidate to know that operation with the generator voltage regulator in AUTO is desired state during grid fluctuations, and to know the setpoints where 230 KV system voltage and 500 KV system voltage drop low enough to require entry into the applicable Technical Specification. RO applicants are required to know entry conditions into Technical Specifications; and the high-voltage 230 KV and 500 KV system minimum voltages for operability are NOT contained in the Technical Specification 3.16 bases, which may have made this question a SRO-only question.
Answer Choice Analysis A. INCORRECT.
B. INCORRECT.
C. CORRECT.
D. INCORRECT.
Supporting References
-Surry procedure 0-AP-10.18, "RESPONSE TO GRID INSTABILITY," rev 8, especially steps 6 and 7 on p. 4-5.
References Provided to Applicant None The High Voltage 230 kV and 500 kV are NOT above the line in TS, so it this RO knowledge? It is unclear why Unit 1 would declare the 500kV system inoperable, this is a Unit 2 component. There appears to be some confusion on the offsite power supplies.
Recommend a replacement question.
Answer: C
- 18. 0000W/E04 EA2.1 2 Unit 1 Initial Conditions:
A loss-of-coolant accident (LOCA) occurred from 100% power.
Operators are implementing 1-ECA-1.2, "LOCA OUTSIDE CONTAINMENT."
Current conditions:
Operators have completed all the steps in ECA-1.2 that (a) attempt to verify proper valve alignment, and (b) locate and isolate the leak.
RCS pressure continues to DECREASE.
Annunciator 1B-F3, SFGS AREA SUMP HI LEVEL, is LIT.
1-VG-RM-110, VENT VENT 2 GAS, is below the HIGH setpoint, but is rapidly trending UP.
RM-GW-130-1, PROCESS VENT STK PART, is below the HIGH setpoint, but is rapidly trending UP.
Based on the current conditions, which ONE of the following:
(1) is the NEXT overall mitigating strategy that should be implemented, as specified by ECA-1.2, AND (2) a filtered release to the environment ___________________ exist?
A. (1) Depressurize S/Gs to inject accumulators.
(2) DOES B. (1) Conserve RWST inventory.
(2) DOES C. (1) Depressurize S/Gs to inject accumulators.
(2) DOES NOT D. (1) Conserve RWST inventory.
(2) DOES NOT K/A LOCA Outside Containment Ability to determine and interpret the following as they apply to the LOCA Outside Containment: Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
(CFR: 43.5 / 45.13) (RO - 3.6)
K/A Match Analysis The question allows applicants to demonstrate their ability to follow procedures by knowing the overall mitigating strategy inherent in ECA-1.2, given an operationally valid situation, and using systems knowledge to identify an abnormal situation (radioactive release). The question is RO-level knowledge because it is testing the applicant's understanding of the overall mitigating strategy of an EOP, as specified by the major action categories of the EOP.
Answer Choice Analysis
A. INCORRECT. (1) The next mitigating strategy/major action category is to transition to ECA-1.1 to conserve RWST inventory. Plausible because depressurizing S/Gs to inject accumulators would rapidly cool-down and depressurize the RCS, as well as assisting the operator in maintaining RCS inventory that is being lost via the LOCA. (2) is the correct option; a release path exists by definition because the LOCA is outside containment; also, the radiation monitors are not yet in alarm and ventilation has not yet isolated.
B. CORRECT. (1) The next step in the ECA-1.2 procedure would be to transition to ECA-1.1 to conserve RWST inventory. This is stated explicitly on p. 6 of the lesson plan for ECA-1.2. (2) is also correct, as described for A. above.
C. INCORRECT. (1) see analysis for A. above; (2) is incorrect regarding the existence of a release path. Plausible if the applicant believes that radiation levels below the alarm setpoint do not count as a release path, or that realignment of ventilation when radiation levels exceed the alarm setpoint will secure (vs. mitigate) the release path.
D. INCORRECT. (1) is correct, (2) incorrect [as detailed above].
Supporting References
- Surry procedure 1-ECA-1.2; complete procedure.
- Westinghouse (WOG) background document for ECA-1.2, rev 2, especially p. 6
- the second part of this question is modified from question W/E04EK1.2 that appeared on the 2006 Surry RO exam.
- Surry lesson plant ND-95.3-LP-21, "ECA-1.2, LOCA OUTSIDE CONTAINMENT," rev. 7, p. 6.
References Provided to Applicant none Why is this a two part question? Does this exam need to be 200 questions?
Answer: B
- 19. 0000W/E05 EA1.1 2 Unit 1 initial conditions:
Loss of Main and Auxiliary Feedwater EOP transition from E-0 to 1-FR-H.1(RESPONSE TO LOSS OF SECONDARY HEAT SINK)
Current plant conditions:
The 1A SG is to be depressurized to establish condensate flow to the SG.
RCS Pressure is 2000 psig and slowly increasing Steam Generator Levels are as follows:
'A' SG Wide Range Level - 42% and slowly decreasing
'B' SG Wide Range Level - 41% and slowly decreasing
'C' SG Wide Range Level - 43% and slowly decreasing Based on the above conditions: (1) why is steam flow limited to 1.0 x 106 Lbm/hr during the depressurization and (2) what actions are directed by 1-FR-H.1 if adequate condensate flow to the 1A SG does not occur with SG pressure = 550 psig?
A. (1) To prevent main steam line isolation (2) Continue to depressurize 'A' SG until condensate flow established B. (1) To prevent main steam line isolation (2) Commence RCS Bleed and Feed C. (1) To limit cooldown rate to < 100°F/Hr (2) Continue to depressurize 'A' SG until condensate flow established D. (1) To limit cooldown rate to < 100°F/Hr (2) Commence RCS Bleed and Feed K/A Inadequate Heat Transfer - Loss of Secondary Heat Sink Ability to operate and / or monitor the following as they apply to the (Loss of Secondary Heat Sink): Components, and functions of control and safety systems including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
K/A Match Analysis Requires knowledge of the operations of system components and how to operate them in such a manner as to avoid inadvertent system activation.
Answer Choice Analysis A. Correct: IAW 1-FR-H.1 note, steam flow during depressurization is limited to 1 x10E6 PPH to prevent Main Steam Line Isolation. If one SG depressurization to 550 psig does not provide adequate flow, continue to depressurize the SG.
B. Incorrect: 1st part is correct. 2nd part is incorrect because Bleed and Feed criteria are not met. 2nd part is plausible because the next step in FH-H.1 if performance of Bleed and Feed if criteria is met.
C. Incorrect: 1st part is plausible because in other portions of the EOP rapid depressurization of the SGs are limited to 100 0F/Hr. 2nd part is correct.
D. Incorrect: 1st part is plausible because in other portions of the EOP rapid depressurization of the SGs are limited to 100 0F/Hr. 2nd part is incorrect because Bleed and Feed criteria are not met. 2nd part is plausible because the next step in FH-H.1 if performance of Bleed and Feed if criteria is met.
Supporting References ND-95.3-LP-41Step 13-18, Obj: B 1-FR-H.1(Response to Loss of Secondary Heat Sink)
References Provided to Applicant none Question altered to maintain primary question and follow procedure flow path in FH-H.1 Answer: A
- 20. 0003A3.04 3 Initial Unit 1 Conditions:
- Power Range NIs are: N-41=36.3%, N-42=36.2%, N-43=34.7%, N-44=36.6%
- Delta T Power is 35%
- All three RCPs are operating Present Unit 1 Conditions:
- RCP 1A frame vibrations indicate 10 mils
- 2 minutes after the frame vibrations increase to 10 mils, the speed sensing panel actuates for RCP 1A.
Based on the above conditions, which one of the following:
(1) correctly states the status of the 1C-H4, RCP FRAME DANGER, annunciator and (2) the impact on reactor operations from the actuation of the speed sensing panel?
A. (1) 1C-H4 is illuminated.
(2) Unit 1 reactor automatically trips.
B. (1) 1C-H4 is illuminated.
(2) Unit 1 reactor does NOT automatically trip.
C. (1) 1C-H4 is NOT illuminated.
(2) Unit 1 reactor automatically trips.
D. (1) 1C-H4 is NOT illuminated.
(2) Unit 1 reactor does NOT automatically trip.
K/A:
003A3.04 Reactor Coolant Pump Ability to monitor automatic operation of the reactor coolant pumps including RCS flow.
K/A MATCH ANALYSIS:
The question requires the applicant to know how the low RCS flow (as caused by an automatic RCP trip) affects the reactor status, thereby testing knowledge of automatic operation of RCPs including RCS low flow.
ANSWER CHOICE ANALYSIS:
A. Correct. With 3/4 PRNIs > 35%, the reactor will trip on low flow. 1C-H4 alarms at 5 mils.
B. Incorrect. Rx will trip with 3/4 PRNIs >35%. Plausible because one PRNI is less B. than 35% and delta-T power is not greater than 35%. Could also be plausible if A applicant has a misconception of the power level at which the low flow trip is
. blocked.
C Incorrect. The second part is incorrect because frame vibes are > 5 mils. Plausible
. because 1C-H5, RCP SHAFT DANGER, alarms at 20 mils. Therefore, if the C applicant confuses the two vibration alarms for an RCP, they would conclude that
. NOT illuminated is correct.
B D Incorrect. See above.
D C
REFERENCES:
1C-H4, RCP FRAME DANGER, Revision 3 1C-H5, RCP SHAFT DANGER, Revision 2 ND-93.3-LP-16, Permissive/Bypass/Trip Status Lights, Revision 9 ND-88.1-LP-6-DRR, Reactor Coolant Pumps, Revision 19 Answer: A
- 21. 0003A4.08 3 A tube leak has developed in the thermal barrier heat exchanger for 1-RC-P-1A.
Once the indicated thermal barrier flow has exceeded 50 gpm, which ONE of the following (1) states the expected automatic plant response and (2) the associated time delay for that response?
A. (1) 1-CC-TV-120A (A RCP Thermal Barrier Isolation Trip Valve) will close (2) 50 seconds.
B. (1) 1-CC-TV-120A (A RCP Thermal Barrier Isolation Trip Valve) will close (2) 10 seconds.
C. (1) 1-CC-TV-140A (Thermal Barrier CC Return Trip Valve) and 1-CC-TV-140B (Thermal Barrier CC Return Trip Valve) will close (2) 50 seconds.
D. (1) 1-CC-TV-140A (Thermal Barrier CC Return Trip Valve) and 1-CC-TV-140B (Thermal Barrier CC Return Trip Valve) will close (2) 10 seconds.
K/A Reactor Coolant Pumps Ability to manually operate and/or monitor in the control room Reactor Coolant Pumps cooling water flow.
K/A Match Analysis Requires applicants to be able to monitor thermal barrier flow during a transient and be able to use this information to determine the impact on RCP cooling water flow.
Answer Choice Analysis (See drawing contained in supporting documentation.)
A. Incorrect CC-TV-120A will close, but it will close after 10 seconds.
Plausible if the candidate confuses the flow setpoint with the time requirement.
B. Correct C. In-Correct - Plausible as the 140 valves will isolate on thermal barrier high flow, but the setpoint is higher. The time is incorrect as well, but plausible as described in distractor A.
D. In-Correct - Plausible as the 140 valves will isolate on thermal barrier high flow, but the setpoint is higher. The time is correct.
Supporting References ND-88.5-LP-1, Component Cooling System, Rev. 23 - Obj C References Provided to Applicant none
Repeat concept of Question 16.
Answer: B
- 22. 0004A4.08 3 Current plant conditions are as follows on Unit 2:
Reactor power is 100%
Pressurizer level is 50% and slowly lowering.
Letdown flow is 104 gpm The team initiated 1-AP-16.00 (Excessive RCS Leakage) and completed the immediate actions of 1-AP-16.00. At this time, the reactor operator reports the following:
Seal Injection Flow to each RCP is 7 gpm Seal Leak-off Flow for each RCP is 3 gpm Tave is stable at 573 °F The leak rate was determined to be 75 gpm Which ONE of the following correctly describes what value the RO should set charging flow at to maintain pressurizer level stable?
A. 54 GPM B. 63 GPM C. 75 GPM D. 84 GPM K/A Chemical and Volume Control System Ability to manually operate and/or monitor in the control room Chemical and Volume Control System: Charging.
K/A Match Analysis Requires applicant to know what to set the charging flow controller for a given leak rate..
Answer Choice Analysis A. In-Correct but plausible if the candidate fails to account for leak-off flow when determining how charging flow should be operated.
B. Correct - 75 gpm (leak rate) + 9 gpm (seal leak-off) - 21(seal injection) = 63
gpm C. In-Correct but plausible if the candidate does not account for seal leak off or seal injection.
D. In-Correct but plausible if the candidate accounts for seal leak-off, but not seal injection.
Supporting References ND-93.3-LP-7, Pressurizer Level Control System, Rev. 09 - Obj D ND-88.3-LP-2, Charging and Letdown, Rev. 025 References Provided to Applicant none Fix Mark numbers for 2-CH-FCV-2122 and Unit 2 Charging Pumps and HCV If at Max output VCT level would be decreasing.
This needs to be checked on the simulator.
Answer: B
- 23. 0005 G2.4.9 3 Unit 1 Conditions:
Cold Shutdown ~ 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after an extended full power run 1-RC-LI-100A = 12.0 ft.
1A RHR pump operating and pump amps are oscillating 1-RH-FI-1605 = 2000 gpm and oscillating The team has increased RCS level back into the acceptable region and secured 1-RH-P-1A in accordance with 1-AP-27.00 (LOSS OF DECAY HEAT REMOVAL CAPABILITY).
Which ONE of the following actions are directed by 1-AP-27.00 to restore RHR flow?
A. Vent the 1A RHR pump, restart the 1A RHR pump and verify the RHR heat sink.
B. Vent the 1B RHR pump, start the 1B RHR pump and verify the RHR heat sink.
C. Close 1-RH-HCV-1758 and 1-RH-FCV-1605, then re-start the 1A RHR pump and throttle to the pre-event rate.
D. Close 1-RH-HCV-1758 and 1-RH-FCV-1605, then start the 1B RHR pump and throttle to the pre-event rate.
Knowledge of low power / shutdown implications in accident (e.g. loss of coolant accident or loss of residual heat removal) mitigation strategies.
K/A Match Analysis Requires knowledge of loss of RHR mitigation strategies Answer Choice Analysis A. Incorrect: Plausible because this would be correct if the 1B RHR pump were not available.
B. Incorrect: Plausible because there is air in the common suction line and there would be some air getting to the 1B RHR pump so it would make sense to vent the 1B RHR pump prior to starting.
C. Incorrect: 1st part is correct. 2nd part is plausible because the 1A RHR pump is restarted if the 1B RHR pump is unavailable.
D. Correct: Correct per AP/27 Supporting References AP/27 Step 11, ND-95.2-LP-12 Obj: D/F 1-OP-RH-001 References Provided to Applicant none Answer: D
- 24. 00068 AG2.4.42 2 Plant conditions:
Unit 1 = 100%
Unit 2 = shutdown for refueling MAIN CONTROL ROOM OXYGEN MONITOR alarms A hazardous substance has been spilled in the main control room 0-AP-20.00, MAIN CONTROL ROOM INACCESSIBILITY, is initiated Operators proceed to the Auxiliary Shutdown Panel with FCA procedures Based on the above conditions, which ONE of the following states equipment that is operated from the Auxiliary Shutdown Panel as directed by 0-AP-20.00?
A. CC pumps
B. RHR pumps C. 1-CH-HCV-1311 (Aux Pressurizer Spray)
D. 1-CH-FCV-1122 (Charging Flow Controller)
K/A Control Room Evac.
Knowledge of emergency response facilities.
K/A Match Analysis Requires knowledge of equipment used during Control Room Evacuation.
Answer Choice Analysis A. Incorrect: Plausible because it is directed by AP/20 to be operated locally if
- needed, but it is not operated from the ASP.
B. Incorrect: Plausible because it is directed by AP/20 to be operated locally if needed, but it is not operated from the ASP.
C. Incorrect: Plausible because charging is controlled from the ASP and Pzr heaters are controlled in Manual (wont de-energize as pressure increases)
D. Correct:
Supporting References AP/20 MAIN CONTROL ROOM INACCESSIBILITY ND-93.4-LP5 Obj: B&D References Provided to Applicant none OK, but K/A issue and correct procedure would be FCA-1.00 vice AP-20.0.
Answer: D
- 25. 0006K1.05 2 Initial plant conditions on Unit 1 are as follows:
Reactor tripped following an inadvertent SI.
Current plant conditions on Unit 1 are as follows:
1-ES-1.1, SI Termination, has been initiated.
SI signal has been reset.
Actions have been completed to re-establish charging and letdown.
Seal leakoff flow from each RCP is ~3 gpm.
PRT level shows a slow rising trend.
VCT level is 43% and decreasing.
Reactor Pressure is stable at 2235 psig.
Tailpipe temperatures are 105°F and stable.
Which ONE of the following would be consistent with the above conditions?
A. 1-CH-MOV-1381 Seal Return Header Isolation Valve is closed causing 1-CH-RV-1382B Seal Return Heat Exchanger Relief Valve to lift.
B. A component cooling water leak has developed on the Seal Return Heat Exchanger causing 1-CH-RV-1382B Seal Return Heat Exchanger Relief Valve to lift.
C. 1-CH-MOV-1381 Seal Return Header Isolation Valve is closed causing 1-CH-RV-1382A Seal Return Relief Valve to lift.
D. A component cooling water leak has developed on the Seal Return Heat Exchanger causing 1-CH-RV-1382A Seal Return Relief Valve to lift.
K/A Emergency Core Cooling System Knowledge of the physical connections and/or cause-effect relationships between Emergency Core Cooling and the following systems: RCP seal injection and return.
K/A Match Analysis Requires applicant to recognize that 1-CH-MOV-1381, Seal Return Header Isolation Valve, closes on a SI (ECCS) actuation signal.
Answer Choice Analysis A. In-Correct but plausible since 1-CH-MOV-1381 closes on an SI actuation and must be manually reopened. However, 1-CH-RV-1382B is downstream of the 1-CH-MOV-1381 and would not be experiencing system pressure. Also 1-CH-RV-1382B relieves to the VCT not the PRT.
B. In-Correct but plausible since component cooling water is at a higher pressure than Seal Water Return pressure and could cause 1-CH-RV-1382B to lift and relieve.
However, a leak on the Heat Exchanger would be indicated by an increasing level in the VCT not the PRT.
C. Correct CH-MOV-1381 closes on an SI actuation and must be manually reopened. As pressure in the seal return line builds, leak-off flow would
decrease until the pressure reached 150 psig causing 1-CH-RV-1382A to lift and relieve to the PRT.
D. In-Correct but plausible since component cooling water is at a higher pressure than Seal Water Return pressure. However, 1-CH-RV-1382A is upstream of 1-CH-MOV-1381 and would not be affected by a component cooling water leak.
Supporting References ND-88.3-LP-3, Seal Injection, Rev. 07 - Obj F References Provided to Applicant none Why wouldn't VCT level be decreasing? 9 gpm not making it back to VCT.
Answer: C
- 26. 0007A1.03 1 Unit 1 Initial Conditions:
100% Power Pressurizer Relief Tank (PRT) level is 62% and STABLE A leaking pressurizer code safety valve has been identified.
Current conditions:
A PRT annunciator has just alarmed PRT level is 76% and INCREASING PRT temperature is 126 °F and INCREASING PRT pressure is 3 psig and INCREASING Based on the current conditions, which one of the following identifies (1) the parameter that caused the PRT annunciator, AND (2) chemistry will sample the PRT gas space when PRT level INCREASES by 5% because large changes in PRT level can result in an undesirable atmosphere in the PRT gas space _________(2)______?
A. (1) High level (2) due to PG O2 off-gassing as PRT temperature changes.
B. (1) High temperature (2) due to PG O2 off-gassing as PRT temperature changes.
C. (1) High level (2) due to H2 concentration potentially exceeding 4% in the process vent.
D. (1) High temperature (2) due to H2 concentration potentially exceeding 4% in the process vent.
K/A System 007: Pressurizer Relief Tank/Quench Tank System (PRTS)
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including: monitoring quench tank temperature.
(CFR: 41.5 / 45.5 ) (RO - 2.6)
K/A Match Analysis Given a plausible scenario that would require monitoring of PRT parameters (leaking safety valve), the candidate must demonstrate ability to recognize the correct abnormal parameter and the correct reason chemistry samples the PRT gas space when PRT level and temperature changes.
Answer Choice Analysis NOTE TO SURRY: Please explain what is meant by "PG O2." We may need to spell this out on the test instead of just symbols (what is in the procedure).
A. INCORRECT. Surry procedure 1-OP-RC-011, "Pressurizer Relief Tank Operations," table in step 5.1.1, identifies high level at 83% and high temperature at 125 °F. Therefore, high temperature is the correct choice. 72% is plausible, because this is greater than 75% (a possible high level setpoint). (2) is the correct choice, taken word-for-word from the precautions & limitations, and notes in section 5.2 of 1-OP-RC-011.
B. CORRECT. See above explanation; high temperature is correct, correct reason for chemistry sample.
C. INCORRECT. (1) is incorrect as per the above. (2) is also incorrect, but plausible. Another precaution and limitation (4.9) in 1-OP-RC-011 states:
"...these limits ensure Hydrogen concentration in the Process Vent remains below the 4% flammability limit."
D. INCORRECT. (1) is correct choice, (2) incorrect as per the above.
Supporting References
- Surry UFSAR Table 4.1-3, "PRESSURIZER AND PRESSURIZER RELIEF TANK DESIGN DATA." REV 36, P. 4.1-21.
-Surry Unit 1 procedure 1-OC-RC-011, "PRESSURIZER RELIEF TANK OPERATIONS." REV 23, p. 7,8,9.
References Provided to Applicant none Why two part question?
K/A relationship of part 2 is unclear. This is two questions.
Answer: B
- 27. 0008 A2.04 2 Unit 1 plant conditions:
Reactor power = 100%
Charging flow = 100 gpm increasing 1-CC-RI-105 (CC Heat Exchanger A/B Outlet Radiation Monitor) alarms HIGH CC surge tank level = 64% increasing 1-AP-16.00 (EXCESSIVE RCS LEAKAGE) is initiated Based on the above conditions, which ONE of the following describes where the excess volume in the CC system will go to and (2) what actions is directed first by 1-AP-16.00 to attempt to isolate the leak?
A. (1) The process vent system (2) Isolate letdown B. (1) The process vent system (2) Isolate thermal barrier on suspected RCP C. (1) The auxiliary building sump (2) Isolate letdown D. (1) The auxiliary building sump (2) Isolate thermal barrier on suspected RCP K/A Component Cooling Water System (CCWS)
Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PRMS alarm.
K/A Match Analysis Requires knowledge of the impact of a PRMS alarm on CCW system operations and how the AP mitigates the event.
Answer Choice Analysis A. Incorrect: 1st part is incorrect because the vent to the process vent system closed when the rad monitor alarmed. 1st part is plausible because if there were no
- alarm, it would be correct. 2nd part is correct per AP/16.
B. Incorrect: 1st part is incorrect because the vent to the process vent system closed when the rad monitor alarmed. 1st part is plausible because if there were no
- alarm, it would be correct. 2nd part is incorrect because AP/16 isolates systems in order of most likely to stop the leak. Letdown is isolated before the RCPs are examined for suspected leakage. 2nd part is plausible because it is performed by AP/16.
C. Correct: When the rad monitor alarms, the CC surge tank vent to the process vent system closes. As pressure increases in the system due to the RCS leak, system pressure increases to 35 psig when the tank relief lifts and discharges to the AB sump. Indications are entry conditions for AP/16 EXCESSIVE RCS LEAKAGE which isolates systems in order of most likely to stop the leak. Letdown is isolated before the RCPs are examined for suspected leakage.
D. Incorrect: 1st part is correct. 2nd part is incorrect because AP/16 isolates systems in order of most likely to stop the leak. Letdown is isolated before the RCPs are examined for suspected leakage. 2nd part is plausible because it is performed by AP/16.
Supporting References ND-93.5-LP1
ND-88.5-LP-1 Obj: A 1-AP-16.00 (Excessive RCS Leakage)
References Provided to Applicant none Fix the following: Initial charing flow value. It is always about 90 gpm.
Need to remove the excess letdown portion and increase surveillance portion.
K/A has two parts, so two parts is OK.
Answer: C
- 28. 0008 G2.1.7 2 Unit 1 Initial Conditions:
75% Power at Middle-of-Life (MOL) conditions.
Rod Control is in AUTOMATIC.
VCT automatic makeup controls are set to the current RCS boron concentration.
Excess Letdown is in service in preparation for removing Normal Letdown from service.
Current conditions:
Component Cooling (CC) surge tank level is slowly DECREASING at 1%
every 5 minutes.
Reactor power is slowly INCREASING.
Based on the current conditions and assuming NO other operator actions, which ONE of the following identifies (1) the location of the CC leak that would cause the current conditions, AND (2) the expected impact of the CC leak on rod control?
LEAKING COMPONENT ROD CONTROL A. (1) Seal Water Heat Exchanger (2) Rods will step IN.
B. (1) Seal Water Heat Exchanger (2) Rods will step OUT.
C. (1) Excess Letdown Heat Exchanger (2) Rods will step IN.
D. (1) Excess Letdown Heat Exchanger (2) Rods will step OUT.
K/A 008 Component Cooling Water (CCW)
Generic 2.1.7: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
(CFR: 41.5 / 43.5 / 45.12 / 45.13 ) (RO - 4.4)
K/A Match Analysis The RO applicant, given an operationally valid condition and a change in reactor behavior, will demonstrate the ability to analyze the condition and predict the operational impacts of the condition.
Answer Choice Analysis A. CORRECT. A leak in the Seal Water HX will leak to the CVCS system, which will result in a dilution and inward rod motion.
B. INCORRECT. Plausible as a leak in the Seal Water HX will leak to the CVCS system which will result in a dilution. Rods will move in not out, but plausible if the candidate believes reactor power has increased due to rod motion out, or if the candidate believes that the nuclear power/turbine power mismatch circuitry will actuate before the Tave/Tref mismatch.
C. INCORRECT. Plausible as Excess Letdown has just been placed in service, but with the pressure in this HX, the RCS will leak to the CCW system. Rods will move in.
D. INCORRECT. Plausible as Excess Letdown has just been placed in service, but with the pressure in this HX, the RCS will leak to the CCW system. Rods will move in not out, but plausible if the candidate believes reactor power has increased due to rod motion out, or if the candidate believes that the nuclear power/turbine power mismatch circuitry will actuate before the Tave/Tref mismatch.
Supporting References
-Surry lesson plan ND-88.3-LP-2, "Charging and Letdown," rev. 15.
References Provided to Applicant none OK.
Answer: A
- 29. 000W/E07 EK2.1 2 In FR-C.2 (Response to Degraded Core Cooling), if attempts to establish adequate
core cooling using the High Head Safety Injection System are ineffective, the intact Steam Generators are depressurized to 200 psig and subsequently to atmospheric pressure.
Which ONE of the following describes (1) the purpose of the depressurization of the steam generators and (2) why the steam generators are depressurization is stopped at 200 psig?
A. (1) To depressurize the RCS to increase accumulator and Low Head Safety Injection Flow.
(2) To isolate the Safety Injection Accumulators and prevent Nitrogen addition into the RCS.
B. (1) To depressurize the RCS to increase accumulator and Low Head Safety Injection Flow.
(2) To prevent the loss of RCP support conditions by maintaining adequate RCS pressure.
C. (1) To maximize Auxiliary Feedwater Flow and enhance RCS cooldown.
(2) To isolate the Safety Injection Accumulators and prevent Nitrogen addition into the RCS.
D. (1) To maximize Auxiliary Feedwater Flow and enhance RCS cooldown.
(2) To prevent the loss of RCP support conditions by maintaining adequate RCS pressure.
K/A Degraded Core Cooling Knowledge of the interrelations between the (Degraded Core Cooling) and the following: components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(CFR: 41.7/45.7) (RO - 3.6)
K/A Match Analysis Requires the applicant to demonstrate knowledge Degraded Core Cooling and manual actions that must take place during the performance of FR-C.2.
Answer Choice Analysis A. CORRECT.
B. INCORRECT. (1) correct. (2) incorrect, but plausible as the goal of this step is to hold RCS pressure, but not for RCP maintenance, but for N2 injection prevention.
C. INCORRECT. (1) incorrect, but plausible if candidate belives that lower SG pressure will result in higher AFW flow, which will remove the degraded core cooling condition. (2) correct.
D. INCORRECT. (1) incorrect, but plausible if candidate belives that lower SG pressure will result in higher AFW flow, which will remove the degraded core cooling condition. (2) incorrect, but plausible as the goal of this step is to hold RCS pressure, but not for RCP maintenance, but for N2 injection prevention.
Supporting References FR-C.2, "RESPONSE TO DEGRADED CORE COOLING,"
References Provided to Applicant none Answer: A
- 30. 000W/E08 EK1.3 3 Concerning a pressurized thermal shock event, a ___(1)____ would cause the most significant Pressurized Thermal Shock (PTS) challenge to the plant, and in accordance with FR-P.1 (Response to Imminent Pressurized Thermal Shock), an appropriate response would be to ____(2)____?
A. (1) Main Steam Line Break (2) Stop the RCS cooldown and reduce Safety Injection flow B. (1) Main Steam Line Break (2) Reduce the RCS cooldown rate to < 100 °F/hr and depressurize the RCS to establish 30 °F subcooling C. (1) Steam Generator Tube Rupture (2) Stop the RCS cooldown and reduce Safety Injection flow D. (1) Steam Generator Tube Rupture (2) Reduce the RCS cooldown rate to < 100 °F/hr and depressurize the RCS to establish 30 °F subcooling K/A RCS Overcooling - PTS / 4 Knowledge of the operational implications of the following concepts as they apply to the (Pressurized Thermal Shock): annunciators and conditions indicating
signals, and remedial actions associated with the (Pressurized thermal Shock).
K/A Match Analysis Requires knowledge of events that result in PTS and actions to be taked to combat the event.
Answer Choice Analysis A. Correct: Part 1 is correct. Part 2 is correct in accordance with FR-P.1 B. Incorrect: Part 1 is correct. 2nd part is plausible as a 100 °F/hr cooldown rate would prevent entry into FR-P.1 and there are actions within FR-P.1 to reduce RCS pressure.
C. Incorrect: 1st part is plausible a Steam Generator Tube Rupture could result in PTS, however, the question asks which is the most significant. Part 2 is correct.
D. Incorrect: 1st part is plausible a Steam Generator Tube Rupture could result in PTS, however, the question asks which is the most significant. 2nd part is plausible as a 100 °F/hr cooldown rate would prevent entry into FR-P.1 and there are actions within FR-P.1 to reduce RCS pressure.
Supporting References 1-FR-P.1 95.3-LP-46 Obj. C References Provided to Applicant none Answer: A
- 31. 000W/E13 EK2.2 1 Unit 1 initial conditions:
Reactor power = 100%
Condenser vacuum = 25" Hg degrading rapidly Current plant conditions:
1-ES-0.1 (REACTOR TRIP RESPONSE) in progress Condenser vacuum = 0" Hg Based on the above conditions, which ONE of the following states (1) the temperature at which the RCS would be maintained with no operator action and (2) what temperature the RCS is directed to be maintained by 1-ES-0.1?
A. (1) 550°F (2) 547°F B. (1) 550°F (2) 535°F C. (1) 556°F (2) 535°F D. (1) 556°F (2) 547°F K/A Steam Generator Over-pressure /4 Knowledge of the interrelations between the (Steam Generator Overpressure) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems and relations between the proper operations of the systems to the operation of the facility.
K/A Match Analysis Requires knowledge of interrelationships of the SG overpressure protection and heat removal from the RCS/SGs.
Answer Choice Analysis A. Correct:SG PORV setpoint = 1035 psig + 15 psi = 1050 psia = 550 0F. Per 1-ES-0.1, monitor RCS temperature stable at or trending to 547 0F.
B. Incorrect: 1st part is correct. 2nd part is plausible because 535 0F is used as criteria in shifting from AFW to Main Feedwater (Step 16).
C. Incorrect: 556 0F would be correct if PORV were failed (556 0F coincides with the setpt of the lowest SG Safety valve). 2nd part is plausible because 535 0F is used as criteria in shifting from AFW to Main Feedwater (Step 16).
D. Incorrect: 556 0F would be correct if PORV were failed (556 0F coincides with the setpt of the lowest SG Safety valve). 2nd part is correct.
Supporting References ND89.1-LP-2 Obj B.
1-ES-0.1 (REACTOR TRIP RESPONSE)
References Provided to Applicant none
Answer: A
- 32. 000W/E16 EA1.1 2 A Large Break LOCA has occurred on Unit 1.
Current conditions:
All plant systems operated as designed.
Steam Generator Levels are 14% Narrow Range and stable in all three Steam Generators.
AFW flow to each Steam Generator is 125 gpm.
Containment Pressure is currently 15 psia after peaking at 47 psia.
Containment High Range Radiation Monitors (1-RM-RI-127 and 128) are currently indicating 1 x 104 after peaking at 4.3 x 106 .
Based on the current conditions, which ONE of the states (1) the status of Heat Sink and (2) why?
A. (1) Steam generator Heat Sink requirements are met.
(2) Adequate Steam generator Inventory.
B. (1) Steam generator Heat Sink requirements are met.
(2) Adequate AFW flow.
C. (1) Steam generator Heat Sink requirements are NOT met.
(2) Inadequate Steam generator Inventory and AFW flow is < the limit of 540 gpm D. (1) Steam generator Heat Sink requirements are NOT met.
(2) Inadequate Steam generator Inventory and AFW flow is < the limit of 450 gpm.
K/A High Containment Radiation Ability to operate and/or monitor the following as they apply to the (High Containment Radiation): components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
(CFR: 41.7/45.5/45.6) (RO - 3.1)
K/A Match Analysis The question requires the candidate to know that adverse conditions (based on High Range Containment Radiation) existed and the subsequent control of safety systems (Heat Sink).
Answer Choice Analysis
A. INCORRECT. Part 1 incorrect. Plausible if the candidate believes that adverse have cleared. Conditions have cleared based on Containment Pressure, but once High Containment Radiation exists, they cannot be exitted until Engineering evaluates the situation. Part 2 incorrect, but plausible as SG levels are adequate for non-adverse conditions.
B. INCORRECT. Part 1 incorrect. Plausible if the candidate believes that adverse have cleared. Conditions have cleared based on Containment Pressure, but once High Containment Radiation exists, they cannot be exitted until Engineering evaluates the situation. Part 2 incorrect, but plausible as SG levels are adequate for non-adverse conditions.
C. INCORRECT. Part 1 is correct. Part 2 is partially correct. AFW portion is incorrect, but plausible if the candidate confuses adverse numbers with AFW flow requirement for non-adverse numbers with RCPs running.
D. CORRECT.
Supporting References
- ND-95.3-LP-26
- ND-95.3-LP-3 Objective B References Provided to Applicant none Answer: D
- 33. 0010 A3.02 4 Unit 1 plant conditions:
Plant runback occurs from 100% to 90%
RCS HI PRESSURE ALARM lit Master Pressure Controller output demand is currently 70%
Based on the above conditions, which ONE of the following states (1) the status of the Pressurizer Spray valves and (2) what the maximum Pressurizer spray flow rate is based on?
A. (1) Full Open (2) To prevent the PORVs from opening during a 10% step load decrease.
B. (1) Full Open (2) To prevent exceeding the capacity of the PORVs during a load rejection from 100% power.
C. (1) Modulated Open (< 100%)
(2) To prevent exceeding the capacity of the PORVs during a load rejection from 100% power.
D. (1) Modulated Open (< 100%)
(2) To prevent the PORVs from opening during a 10% step load decrease.
K/A Pressurizer Pressure Control Ability to monitor automatic operation of the Pzr PCS, including: PZR pressure.
K/A Match Analysis Requires knowledge of automatic operation of the Pzr pressure control system and the bases for spray flow.
Answer Choice Analysis A. Correct: Pressurizer High Pressure alarm setpt = 2310 psig and when the reference signal setpt = 2235, the spray valves will be full open @ 2305 psig. Design of the Pzr spray valves is to prevent the PORVs from opening during a 10% step load decrease.
B. Incorrect: 1st part is correct. 2nd part is plausible because the PORVs will open on a load rejection from 100%.
C. Incorrect: 1st part is plausible because if the reference setpoint signal were
- higher, it could be correct. 2nd part is plausible because the PORVs will open on a load rejection from 100%.
D. Incorrect: 1st part is plausible because if the reference setpoint signal were
- higher, it could be correct. 2nd part is correct.
Supporting References ND-93.3-LP-5 Obj B References Provided to Applicant none
Answer: A
- 34. 0012 A4.07 1 Unit 1 plant conditions:
Reactor power = 50% power Reactor Protection System testing in progress
'A' Reactor Trip Breaker is CLOSED
'A' Reactor Trip BYPASS Breaker is CLOSED
'B' Reactor Trip Breaker is CLOSED
'B' Reactor Trip BYPASS Breaker is OPEN Based on the above conditions, if an operator closes (attempts to close) 'B' Reactor Trip BYPASS Breaker which ONE of the following states (1) the status of the Reactor Trip BYPASS Breakers and (2) the status of the Reactor?
A. (1) Both Reactor Trip BYPASS Breakers will open (2) The reactor will trip B. (1) Both Reactor Trip BYPASS Breakers will open (2) The reactor will NOT trip C. (1) Only 'B' Reactor Trip BYPASS Breaker will open (2) The reactor will trip D. (1) Only 'B' Reactor Trip BYPASS Breaker will open (2) The reactor will NOT trip K/A Reactor Protection Ability to manually operate and/or monitor in the control room: M/G set breakers.
K/A Match Analysis Requires knowledge of the manual operation of the RPS MG Set Breakers.
Answer Choice Analysis A. Incorrect: 1st part is correct. 2nd part is plausible because if RTA were open, it would be correct.
B. Correct: If both bypass breakers are closed at the same time, each bypass breaker's trip coil will be energized, and both bypass breakers will open. As long as both RTA and RTB are closed initially, there is no signal that will cause them to open.
C. Incorrect: 1st part is plausible because the breakers are interlocked such that
both breakers can not be closed at the same time. 2nd part is plausible because if RTA were open, it would be correct.
D. Incorrect: 1st part is plausible because the breakers are interlocked such that both breakers can not be closed at the same time.
Supporting References ND-93.3-LP10 Obj: B References Provided to Applicant none Easy, but not a K/A match. Recommend replacement with a K/A match.
Answer: B
- 35. 0012 K1.05 2 Unit 1 initial conditions:
- Reactor power = 5%
- Pressurizer Pressure Protection transmitter (1-RC-PT-1456) failed low (I&C investigating)
Current plant conditions:
- Pressurizer Pressure Protection transmitter (1-RC-PT-1455) subsequently failed low Based on the current plant conditions, which ONE of the following correctly describes (1) the effect on the reactor and (2) the effect these failures on subcooling indication on the Inadequate Core Cooling Monitor (ICCM)?
A. (1) A Reactor Trip will occur (2) No impact on subcooling indication B. (1) A Reactor Trip will NOT occur (2) Subcooling will indicate -35 °F C. (1) A Reactor Trip will occur (2) Subcooling will indicate -35 °F D. (1) A Reactor Trip will NOT occur (2) No impact on subcooling indication
K/A Reactor Protection:
Knowledge of the physical connections and/or cause-effect relationships between the RPS and the following systems: ESFAS K/A Match Analysis Requires knowledge of how an ES initiation (SI) affects the RPS system.
Answer Choice Analysis A. Correct: SI will occur and cause a reactor trip. Subcooling is not affected.
B. Incorrect: Part 1 is plausible because the Low Pzr Pressure reactor trip is not in effect at 5%. Part 2 is plausible if the candidate believes pressure protection feeds the ICCM.
C. Incorrect: 1st part is correct. Part 2 is plausible if the candidate believes pressure protection feeds the ICCM.
D. Incorrect: Part 1 is plausible because the Low Pzr Pressure reactor trip is not in effect at 5%. Part 2 is correct.
Supporting References ND-93.3-LP10 Obj: C, ND-88.1-LP-3 Obj: B, ND-93.3-LP-5 Obj: B NE-93.3-LP-11 References Provided to Applicant none Answer: A
- 36. 0013 K2.01 2 A loss of which ONE of the following busses will result in an imminent Safety Injection?
A. Vital Bus III B. Vital Bus I C. 'A' DC Bus D. 'B' DC Bus
K/A ESFAS:
Knowledge of bus power supplies to the following: ESFAS/safeguards equipment control.
K/A Match Analysis Requires applicant to know that although all choices feed instrumentation and/or control features of the ESFAS, only ONE of the choices will result in an imminent SI.
Answer Choice Analysis A. Correct - A loss of Vital Bus III will result in a channel three high steam flow.
This combined with the reactor trip and subsequent temperature drop will cause an SI.
B. In-Correct but plausible if the applicant believes that Vital Bus I will result in a high steam flow indication issue combined with the low Tave that will occur on the trip. Additionally, the applicant may believe that Vital Bus 1 feeds 'A' Train of SI and that Vital Bus II feeds 'B' Train of SI and that a loss of either will result in an SI. Note that AP-10.03 states that SI is expected.
C. In-Correct but plausible as 'A' DC bus provides control power to the SI system. The applicant may believe that SI is a de-energize to function rather than an energize to function.
D. In-Correct but plausible as 'B' DC bus as this failure will result in a failure of the generator output breakers to open, which could result in an excessive RCS cooldown and subsequent SI.
Supporting References ND-90.3-LP-6 ND-90.3-LP-7 References Provided to Applicant none Answer: A
- 37. 0022 K1.04 2 Unit 2 Initial Conditions:
100% Power.
Chilled CC is in service to Containment.
Current conditions:
2-CD-REF-1, Unit 2 Turbine Building Chiller Unit, trips due to a fault.
Based on the current conditions, which ONE of the following describes (1) the effect on Unit 2 containment indicated partial pressure, AND (2) Unit 2 containment temperature?
A. (1) Indicated partial pressure will INCREASE.
(2) Containment temperature will INCREASE.
B. (1) Indicated partial pressure will DECREASE.
(2) Containment temperature will INCREASE.
C. (1) Indicated partial pressure will INCREASE.
(2) Containment temperature will DECREASE.
D. (1) Indicated partial pressure will DECREASE.
(2) Containment temperature will DECREASE.
K/A 022 Containment Cooling System (CCS)
Knowledge of the physical connections and/or cause-effect relationships between CCS and the following systems: Chilled Water.
(CFR 41.2 to 41.9 / 45.7 to 45.8) (RO - 2.9*)
K/A Match Analysis The question requires the RO applicant to understand the relationship between the failed chiller unit and the effect on containment cooling system.
Answer Choice Analysis A. INCORRECT. (1) Based on the given conditions, containment partial pressure will DECREASE due to the loss of chilled CC. Containment temperature will INCREASE due to the loss of chilled CC. Therefore, for this distractor (1) is incorrect and (2) is correct. All other distractors are plausible logical combinations of the correct trends.
B. CORRECT. See above analyses.
C. INCORRECT. See analyses for A. above.
D. INCORRECT. See analyses for A. above.
Supporting References
-Old Surry exam question 022K3.02 from 2004-301 exam. Modified to re-arrange logical choices.
-Surry lesson plan ND-88.5-LP-1, rev. 23, especially p. 26-27.
References Provided to Applicant none Corrected mark number and name of unit is acceptable.
Answer: B
- 38. 0026 K3.02 3 Unit 1 plant conditions:
Reactor power = 100%
Current conditions:
LBLOCA occurs Containment pressure = 25 psia increasing 1-CS-MOV-101A (Containment Spray Pump 'A' Discharge Valve) does not open 1-CS-MOV-101B (Containment Spray Pump 'A' Discharge Valve) does not open Based on the above conditions, which ONE of the following states:
(1) Which Recirculation Spray (RS) System pump suction(s) is/are being supplied by B Train of Containment Spray?
(2) If sufficient containment spray flow is being supplied to meet the design basis of the CS system?
A. (1) RS Train B ONLY (2) The CS design basis is being met.
B. (1) RS Train B ONLY (2) The CS design basis is NOT being met.
C. (1) RS Train A and B (2) The CS design basis is being met.
D. (1) RS Train A and B (2) The CS design basis is NOT being met.
K/A Containment Spray Knowledge of the effect that a loss or malfunction of the CSS will have on the following: Recirculation spray system.
K/A Match Analysis Requires knowledge of the effect of CS failure on containment recurculating spray system.
Answer Choice Analysis A. Correct: CS train A supplys RS train A ONLY. Only one CS train is required per system purpose.
B. Incorrect: 1st part is correct. 2nd part is plausible because 1 train of CS is not supplying spray to containment.
C. Incorrect: CS train A supplys RS train A ONLY. 1st part is plausible because CS pump discharges can be cross connected but are upstream of the discharge valves.
2nd part is correct.
D. Incorrect: CS train A supplys RS train A ONLY. 1st part is plausible because CS pump discharges can be cross connected but are upstream of the discharge valves. 2nd part is plausible because 1 train of CS is not supplying spray to containment.
Supporting References ND-91-LP-5 Obj: B TS 3.4 Spray Systems References Provided to Applicant none Testing of 'A' Train has no impact on answer and circuit will automatically come out of test.
Pressure changed to absolute to match plant indications Separated 101A and 101B for clarity.
Answer: A
- 39. 0036AA2.01 2 Unit 1 Initial Conditions:
Unit is shutdown for refueling operations.
1-RM-RM-159/160, Containment Particulate/Gas, both read 875 cps.
1-RM-RM-152, New Fuel Storage Area, reads 1.7 mr/hr.
Current conditions:
Core off-load is in progress, when an event occurs.
1-RM-RM-159/160, Containment Particulate/Gas, both read 880 cps.
1-RM-RM-152, New Fuel Storage Area, reads 15.3 mr/hr.
Based on the current conditions, which ONE of the following describes REQUIRED operator actions, in accordance with 0-AP-22.00, "FUEL HANDLING ABNORMAL CONDITIONS?"
(1) Fuel handling operations MUST STOP ____________________,
AND (2) after dumping a train of MCR air bottles in accordance with 0-AP-22.00, THEN
_________________ ?
A. (1) in the Fuel Building. Fuel Handling operations may continue in Containment.
(2) IMMEDIATELY start one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred)
B. (1) in the Fuel Building. Fuel Handling operations may continue in Containment.
(2) Wait 50 minutes before starting one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred)
C. (1) in BOTH the Fuel Building AND Containment (2) Wait 50 minutes before starting one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred)
D. (1) in BOTH the Fuel Building AND Containment (2) IMMEDIATELY start one emergency supply fan (1-VS-F-41 or 2-VS-F-41 preferred)
K/A Fuel Handling Incidents Ability to determine and interpret the following as they apply to the Fuel Handling Incidents: ARM system indications.
(CFR: 43.5/45.13) (RO - 3.2)
K/A Match Analysis The question requires the RO applicant to demonstrate knowledge of how to interpret ARM system readings, as well as utilizing knowledge of the overall mitigating strategy of the AP for fuel handling incidents.
Answer Choice Analysis A. INCORRECT. (1) it is relatively easy to determine that the fuel handling problem is in the Fuel Building based on the ARM indications, harder to discredit that the problem is not in containment because the containment radiation readings in the current conditions are slightly higher than the initial readings.
However, 0-AP-22.00 specifically states that all fuel handling operations must STOP. (2) is incorrect; the procedure specifically requires setting a 50 minute timer before starting the emergency supply fan. Plausible because the 0-AP-22.00 does contain several time-critical operator actions, and the wording is taken directly from step 15 of 0-AP-22.00.
B. INCORRECT. (1) is incorrect, (2) is correct. See analysis of A. above.
C. CORRECT. (1) is correct, (2) is correct. See analysis of A. above.
D. INCORRECT. (1) is correct, (2) incorrect. See analysis of A. above.
Supporting References
- SPS Lesson Plan ND-92.5-LP-7, "Refueling Abnormal Procedures," rev 13, p.
3, 8, and 9.
- Procedure 0-AP-22.00, "FUEL HANDLING ABNORMAL CONDITIONS," rev 22,
- p. 2.
-Modified from Oconee 2009-301 RO exam, question 20.
References Provided to Applicant none MCR emergency ventilation - should say air bottles.
K/A Answer: C
- 40. 0037AG2.4.4 2
Unit 1 is at 20% power during a power increase following a maintenance shutdown.
Initial conditions:
Time = 1000 An existing 2 gallon per day tube leak exists on the 1A SG CHG LINE FLOW = 97 gpm and increasing 1-AP-16.00 (EXCESSIVE RCS LEAKAGE) is entered Current conditions:
Time = 1015 Main Steam Line Rad Monitor level increasing (NOT in alarm)
The NEW Steam Generator Tube leak rate is determined to be 6 gpm Based on the above conditions, which ONE of the following: (1) correctly states if 1-AP-24.00 (MINOR SG TUBE LEAK) is required to be initiated IAW 1-AP-16.00 and (2) what procedure is required to shut down the unit?
A. (1) Yes (2) 1-AP-23.00 (RAPID LOAD REDUCTION).
B. (1) Yes (2) 1-GOP-2.2 (UNIT SHUTDOWN, LESS THAN 30% TO HSD).
C. (1) No (2) 1-AP-23.00 (RAPID LOAD REDUCTION).
D. (1) No (2) 1-GOP-2.2 (UNIT SHUTDOWN, LESS THAN 30% TO HSD).
K/A Steam Generator tube Leak Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.
K/A Match Analysis Requires knowledge of AP entry conditions and Major mitigation strategies.
Answer Choice Analysis A. Correct: From AP/16, step 7, check if radiation levels normal or stable if pre-existing SGTL. RNO is Initiate AP/24. Step 1 in AP/24 if Reactor Trip is not required (no reason it should), Initiate a shutdown per AP/23.
B. Incorrect: 1st part is correct. 2nd part is plausible because if remaining in AP/16
with a tube leak of < 10 gpm, shutdown using the GOP is required.
C. Incorrect: 1st part is plausible because the plant is already operating with a small SGTL. 2nd part is correct if AP/24 is initiated.
D. Incorrect: 1st part is plausible because the plant is already operating with a small SGTL. 2nd part is plausible because if remaining in AP/16 with a tube leak of
< 10 gpm, shutdown using the GOP is required.
Supporting References AP/16 AP/24 References Provided to Applicant none Fix charging flow.
Similar concept to SRO 3, 8 an 19 Answer: A
- 41. 0039 A1.05 1 Unit 1 Initial Conditions:
A Steam Generator (S/G) tube rupture caused an automatic reactor trip and SI from 100% power.
Control room operators are implementing 1-ECA-3.1, "SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY."
Current conditions:
A maximum-rate cooldown in accordance with ECA-3.1 was commenced at time 1500.
The following data has been logged over the last hour: (consider that the time is currently 1600)
TIME RCS COLD LEG TEMP 1500 395 °F 1515 370 °F 1530 346 °F
1545 321 °F 1600 296 °F Based on the current conditions, which ONE of the following correctly describes the cooldown from 1500 to 1600?
(1) The Technical Specification cooldown rate limit ______________________ ,
AND (2) The cooldown is _________________________ .
A. WAS exceeded.
required to be temporarily stopped.
B. WAS exceeded.
NOT required to be stopped.
C. was NOT exceeded.
NOT required to be stopped.
D. was NOT exceeded.
required to be temporarily stopped.
K/A Main and Reheat Steam System (MRSS)
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: RCS T-ave.
(CFR: 41.5/45.5) (RO - 3.2*)
K/A Match Analysis Requires the applicant to demonstrate knowledge of RCS temperature limits on cooldown rates, as related to Tech Specs and an operationally valid ECA-3.1 scenario.
Answer Choice Analysis A. INCORRECT. Tech Spec cooldown rate limit is 100 degrees F in an hour.
ECA-3.1 cooldown rate limit is the same, step 16.a states "maintain cooldown rate in RCS cold legs - LESS THAN 100 degrees F/hr." Therefore, neither the Tech Spec nor the ECA-3.1 cooldown rate was violated in the period from 1500-1600. Distractors are plausible if the applicant believed that ECA-3.1 allowed for a max rate cooldown in excess of Tech Specs (allowed in E-3, but NOT in ECA-3.1), or if the applicant believes that the change over one of the 15-minute, or 30-
minute periods counts toward the 100 F/hr limit.
B. INCORRECT. see analysis for A. above. Plausible if the applicant believes that the 100 degree F/hr limit was violated in either the initial 15 minute period or the last 30 minute period, but the cooldown can continue.
C. CORRECT. see analysis for A. above.
D. INCORRECT. See analysis for A. above. Plausible if the applicant believes that the 100 degree F/hr limit was violated in either the initial 15 minute period or the last 30 minute period, and the cooldown needs to stop.
Supporting References
-steam tables
-Surry procedure 1-ECA-3.1, "SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY," rev. 36, p. 13.
-modified from Indian Point 3 Audit exam question #77 from October 2006.
References Provided to Applicant none Answer: C
- 42. 0039 K4.04 2 Unit 1 Initial Conditions:
A reactor startup is in progress.
All Steam Generator (S/G) Power Operated Relief Valves (PORV) are 10% open, and the controllers are being operated in automatic.
Current conditions:
The 1A S/G PORV controller setpoint is at 1000 psig AND the setpoint is continuously DECREASING.
Based on the current conditions, which ONE of the following correctly describes the effect of this failure if no operator action is taken?
A. If uncorrected, the 522 °F minimum temperature for criticality Tech Spec limit may be violated.
B. If uncorrected, the 545 °F minimum temperature for criticality Tech Spec limit may be violated.
C. 1B and 1C S/G PORVs will open to relieve more steam and maintain Tave within acceptable limits.
D. 1B and 1C S/G PORVs will not respond, and Tave will increase above the limit of 577 °F.
K/A 039 Main and Reheat Steam System (MRSS)
Knowledge of MRSS design feature(s) and/or interlocks which provide for the following: Utilization of steam pressure program control when steam dumping through atmospheric relief/dump valves, including T-ave. limits.
(CFR: 41.7) (RO - 2.9)
K/A Match Analysis RO applicant must demonstrate knowledge of S/G PORV operation in automatic pressure control, along with knowledge of Tave limits.
Question is C/A because the applicant must recall how the S/G PORV controller operates in automatic. Steam header pressure is compared to setpoint. If the setpoint lowers, the S/G PORV will open, releasing more steam. The energy release will cool the RCS, lowering Tave. If uncorrected, the minimum temperature for criticality Tech Spec limit may be violated.
Answer Choice Analysis A. CORRECT. See the above C/A analysis.
B. INCORRECT. Incorrect because the minimum Tech Spec. limit for criticality is 522, not 545. Plausible because the PORV will open, lowering Tave, and because 545 is the low Tave alarm setpoint.
C. INCORRECT. Incorrect because the PORV will open, not close. Plausible because the 1B and 1C S/G PORVs will open to relieve more steam. Also plausible if the candidate confuses the setpoint drifting low with the controller output signal drifting low--which would cause the 1A S/G PORV to go closed, or if the applicant incorrectly picks the wrong direction of valve movement given the setpoint vs. actual pressure difference.
D. INCORRECT. Incorrect because the S/G PORV will open, not close.
Plausible because the high-Tave alarm setpoint is 577 degrees F. Also plausible if the candidate confuses the setpoint drifting low with the controller output signal drifting low--which would cause the 1A S/G PORV to go closed, or if the applicant
incorrectly picks the wrong direction of valve movement given the setpoint vs.
actual pressure difference.
Supporting References
-modified from Turkey Point 2009 exam question #43.
-Surry Technical Specifications 3.1.E, "Minimum Temperature for Criticality," p.
TS 3.1-18.
-Surry ARP for 1H-A3, "HI - LO T AVG LOOP 1A," rev. 4, p. 3.
References Provided to Applicant none There is no correct answer. 1000 psig is 545 °F.
Recommend a new question. Why is the TS included in this question?
Answer: A
- 43. 0059 K4.16 3 Unit 2 plant conditions:
2-GOP 1.5 (UNIT STARTUP, 2% REACTOR POWER TO MAX ALLOWABLE POWER) is in progress Reactor Power = 12%
2-FW-P-1A ('A' Main Feed Water Pump) is feeding all SGs at 1500 gpm each Based on the above plant conditions, which ONE of the following conditions will cause 2-FW-P-1A to trip?
A. MFW pump suction header pressure 65 psig for > 15 sec B. Bus voltage dips to 65% and returns to normal C. A Main Feed Pump Recirc Valve closed for > 15 sec D. 2-FW-P-1A lube oil pressure decreases to 5 psig K/A Main Feedwater Knowledge of MFW design feature(s) and/or interlock(s) which provide for the following: Automatic trips for MFW pumps.
K/A Match Analysis Requires knowledge of MFW pump trip/interocks.
Answer Choice Analysis A. Incorrect: Plausible because suction pressure of > 100 psig is a MF pump start permissive.
B. Correct: Bus voltage ~ 70 % is a MF pump trip C. Incorrect: Plausible because for Unit 1, it would be correct.
D. Incorrect: Plausible because the main turbine low lube oil pressure trip is 6 psig.
Supporting References ND-89.3-LP-3 Obj: E GOP 2.8 References Provided to Applicant none check with utility for 3300 gpm = 1.6 E6 lbm/hr = 12% power Answer: B
- 44. 0060AK3.02 2 Unit 1 plant conditions:
Reactor power = 50%
1-GW-RM-130A (Process Vent Particulate Radiation Monitor) high alarm sounds Based on the above conditions: (1) which ONE of the following valves are interlocked to reposition upon receiving the high alarm and (2) why does that action occur?
A. (1) GW-FCV-101 (Waste Gas Decay Tank bleed FCV)
(2) To isolate a potential release path B. (1) GW-FCV-101 (Waste Gas Decay Tank bleed FCV)
(2) To redirect flow through the waste gas charcoal filters C. (1) GW-FCV-100 (Process Vent Flow Control Valve) (2) To isolate all potential release paths D. (1) GW-FCV-100 (Process Vent Flow Control Valve) (2) To redirect flow through the waste gas charcoal filters
K/A Accidental Gaseous Radwaste Release: Knowledge of the reasons for the following responses as they apply to the Accidental Gaseous Radwaste:
Isolation of the auxiliary building ventilation.
K/A Match Analysis Requires knowledge of the Aux Building Process Ventilation system interlocks and how they act to isolate the system from the unit vent.
Answer Choice Analysis A. Correct: 1-GW-RM-130 (MGPI) high alarm causes GW-FCV-101 (Waste Gas Decay Tank bleed FCV) to close, isolating the WGDTs from the Process Vent system.
B. Incorrect: 1st part is correct. 2nd part is incorrect because it does not redirect any flow, it isolates the tank. 2nd part is plausible because a gas release will normally go through the Waste Gas Charcoal Filters.
C. Incorrect: 1st part is incorrect because GW-FCV-100 (Process Vent Flow Control Valve) is not controlled by the RM in alarm. 1st part is plausible because the rad monitor is down stream of GW-FCV-101 and will monitor all process vent system streams going to the unit vent. 2nd part is incorrect becasue the valve doesn't go closed. 2nd part is plausible because if it did go closed, it would isolate all of the potential release paths from that system.
D. Incorrect: 1st part is incorrect because GW-FCV-100 (Process Vent Flow Control Valve) is not controlled by the RM in alarm. 1st part is plausible because the rad monitor is down stream of GW-FCV-101 and will monitor all process vent system streams going to the unit vent. 2nd part is incorrect because it does not redirect any flow, it isolates the system. 2nd part is plausible because a gas release will normally go through the Waste Gas Charcoal Filters.
Supporting References ND-92.4-LP-1 Obj: B
References Provided to Applicant none Answer: A
- 45. 0061 K5.02 2 Unit 1 Initial Conditions:
The plant operated continuously at 100% power for a period of time before the team manually tripped the reactor due to turbine high vibrations.
All plant systems and component operated as designed.
Current conditions:
Offsite power is NORMAL.
Both motor-driven Auxiliary Feedwater (AFW) pumps 1-FW-P-3A and 1-FW-P-3B are running.
All Steam Generator (S/G) narrow range levels are LESS THAN 12%.
Control room operators have transitioned to 1-ES-0.1, "REACTOR TRIP RESPONSE."
Based on the current conditions, which ONE of the following:
(1) identifies the core burnup at time of the trip that will result in the GREATER required AFW system flowrate to maintain S/G levels stable, AND (2) is the MINIMUM AFW flowrate required by ES-0.1 for the current plant conditions?
CORE BURNUP MINIMUM REQUIRED AFW FLOWRATE A. (1) 1,000 MWD/MTU (2) 350 GPM B. (1) 1,000 MWD/MTU (2) 540 GPM C. (1) 10,000 MWD/MTU (2) 350 GPM D. (1) 10,000 MWD/MTU (2) 540 GPM K/A
061 Auxiliary/Emergency Feedwater (AFW) System Knowledge of the operational implications of the following concepts as they apply to the AFW: decay heat sources and magnitude.
(CFR: 41.5/45.7) (RO - 3.2)
K/A Match Analysis Given a plausible operational situation, requires the applicant to demonstrate knowledge of the relative magnitude and behavior of decay heat sources, as they apply to the AFW system operation. The applicant will also demonstrate knowledge of AFW flowrate requirements as specified in 1-ES-0.1.
Answer Choice Analysis A. INCORRECT. (1) 100% power at a lower core burnup will result in less decay heat load. Plausible if the candidate considers that there is more fuel present in the core at lower burnups, or if the candidate considers that there are certain casualties (such as a main steam line break) that are more severe at lower powers. (2) is also incorrect; ES-0.1 step 2.e) requires greater than 540 gpm AFW flow when RCPs are running and all S/G NR levels are below 12%.
Choice (2) is plausible because 350 gpm would be correct if the RCPs were not running.
B. INCORRECT. (1) is incorrect choice, (2) is correct. See analysis of A. above.
C. INCORRECT. (1) is correct choice, (2) is incorrect. See analysis of A. above.
D. CORRECT. (1) and (2) are correct based on the explanation of A. above.
Supporting References ES-0.1, "REACTOR TRIP RESPONSE," rev. 44, p. 3.
-Question is modified from VC Summer 2007-301 ILO exam, RO question 061K5.02.
References Provided to Applicant none Need to indicate the status of RCPs.
Answer: D
- 46. 0061 K6.01 2 Unit 2 Initial Conditions:
100% Power.
The MCR Undervoltage (UV) bypass switches for turbine-driven Auxiliary Feedwater (AFW) pump 2-FW-P-2 steam supply valves 2-MS-PCV-202A and 2-MS-PCV-202B are in "BYPASS" position for routine maintenance.
At the completion of maintenance, the UV bypass switches are returned to the "NORMAL" position, as required.
Due to an electrical failure, all signals will respond as if the UV bypass switch was still in the "BYPASS" position. Operators are unaware of this condition.
Current conditions:
A loss of all offsite power has occurred.
Safety Injection (SI) is NOT actuated.
Based on the current conditions, which ONE of the following correctly identifies the expected plant response?
A. (1) 2-MS-PCV-202A and -202B will receive automatic open signals.
(2) AFW discharge valves 2-FW-MOV-251A through -251F will receive automatic open signals.
B. (1) 2-MS-PCV-202A and -202B will receive automatic open signals.
(2) AFW discharge valves 2-FW-MOV-251A through -251F will NOT receive automatic open signals.
C. (1) 2-MS-PCV-202A and -202B will NOT receive automatic open signals.
(2) AFW discharge valves 2-FW-MOV-251A through -251F will receive automatic open signals.
D. (1) 2-MS-PCV-202A and -202B will NOT receive automatic open signals.
(2) AFW discharge valves 2-FW-MOV-251A through -251F will NOT receive automatic open signals.
K/A 061 Auxiliary / Emergency Feedwater (AFW) System Knowledge of the effect of a loss or malfunction of the following will have on the AFW components: Controllers and positioners.
(CFR: 41.7 / 45.7)
K/A Match Analysis RO applicants will demonstrate knowledge of the operational impacts of a
malfunction on an extremely risk-important AFW system.
Answer Choice Analysis SURRY: PLEASE CAFEFULLY VALIDATE THE CORRECT ANSWER FOR TECHNICAL ACCURACY. IF THE MOVS GET AUTO-OPEN SIGNALS FROM S/G LO-LO LEVEL CONDITIONS THAT ARE NOT BLOCKED BY THE UV BYPASS SWITCH, CORRECT ANSWER WILL CHANGE TO 'A' A. CORRECT. Both recieve auto open signals from SG LO LO Level.
B. INCORRECT. Part 1 correct. Part 2 incorrect if the candidate fails to identify that a LO LO level will exist and focuses solely on the UV defeat.
C. INCORRECT. see analysis of B. above.
D. INCORRECT. see analysis of B. above.
Supporting References
-Surry lesson plan ND-89.3-LP-4, "AUXILIARY FEEDWATER," rev. 25, p. 13-14.
References Provided to Applicant none This question is difficult, but meets the K/A. It does need to be modified.
Answer: A
- 47. 0062 K2.01 1 Unit 1 Initial Conditions:
100% Power.
Current conditions:
Loss of letdown.
Steam dump control NOT affected.
Loss of Component Cooling to ALL RCP thermal barrier heat exchangers.
Component Cooling to ALL other RCP heat exchangers is NOT affected.
Based on the current conditions, which ONE of the following correctly identifies the vital AC bus or buses that has/have been de-energized?
A. Vital Bus I is de-energized. Vital Buses II, III, and IV are energized.
B. Vital Bus I and Vital Bus III are de-energized. Vital Buses II and IV are energized.
C. Vital Bus II is de-energized. Vital Buses I, III, and IV are energized.
D. Vital Bus II and Vital Bus IV are de-energized. Vital Buses I and III are energized.
K/A A.C. Electrical Distribution Knowledge of bus power supplies to the following: Major system loads.
(CFR: 41.7) (RO - 3.3)
K/A Match Analysis Given a plausible operational scenario, the RO applicant must demonstrate an application of his or her knowledge of vital bus loads to diagnose the abnormal plant condition.
Answer Choice Analysis A. CORRECT. All indications come from a loss of vital bus I only. All other distractors are plausible, but incorrect, combinations of the other vital buses.
B. INCORRECT.
C. INCORRECT.
D. INCORRECT.
Supporting References
- Surry lesson plan ND-90.3-LP-5, "VITAL AND SEMI-VITAL BUS DISTRIBUTION," rev. 15, p. 15-16.
-this question is taken directly out of the Surry ILO exam bank, question ID:
EE00013 on p. 66 of 786.
References Provided to Applicant none Answer: A
- 48. 0062 K3.03 1 Unit 1 plant conditions:
Reactor power = 100%
UPS 1A1 Battery charger fails Based on the above conditions, which ONE of the following actions will automatically occur to power loads on DC bus 1A?
A. UPS 1B1 will power DC bus 1A B. Battery 1A will pickup loads for the next two hours C. DC bus 1A will cross connect to DC bus 1B D. UPS 1A2 will power DC bus 1A K/A AC electrical Distribution.
Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following: DC system.
K/A Match Analysis Requires knowledge of how the malfunction of an AC system will have on DC systems.
Answer Choice Analysis A. Incorrect: Plausible because it could power DC Bus 1A through DC Bus 1B if that line up were directed.
B. Incorrect: Plausible because this would be true if both AC chargers were lost.
C. Incorrect: Plausible because this would be a manual action if DC bus 1A power were lost.
D. Correct: 2 UPS normally supply the DC buses.
Supporting References 90.3-LP-6 Obj: A
References Provided to Applicant none Answer: D
- 49. 0063 K3.02 2 A loss of 'A' DC Bus occurs followed by a Safety Injection. Which ONE of the following is correct regarding the operation of 1-SI-P-1A ('A' Low Head Safety Injection Pump)?
A. 1-SI-P-1A is NOT running but can be started from the MCR.
B. 1-SI-P-1A is NOT running and can NOT be started from the MCR.
C. 1-SI-P-1A is running and can be stopped from the MCR.
D. 1-SI-P-1A is running but can NOT be stopped from the MCR.
K/A DC Electrical Distribution.
Knowledge of the effect that a loss or malfunction of the DC Electrical System will have on the following: Components using DC control power.
K/A Match Analysis Requires the applicant to know the major breakers supplied control power from 1A DC Bus.
Answer Choice Analysis A. In-Correct but plausible since the 'A' LHSI pump will not be running. It cannot be started from the MCR. Plausible if the candidate believes that all 480 V components utilize internal power for control power and forgets that LCC 480V components utilize DC power.
B. Correct C. In-Correct but plausible see distactor 'A'.
D. In-Correct but plausible if the candidate believes that since the charging springs are charged and one breaker operation is normally permitted the breaker will close, but the loss of DC power will prevent opening the breaker.
Supporting References ND-90.3-LP-6, 125 VDC Distribution, Rev. 018, Obj. D References Provided to Applicant none
Answer: B
- 50. 0064 K4.10 2 Unit 1 Initial Conditions:
A spurious safety injection from 100% power occurred four (4) minutes ago.
Current conditions:
An electrical grid transient has JUST resulted in a Station Blackout.
Based on the current conditions, which ONE of the following correctly identifies the SEQUENCE that ALL equipment will automatically load onto the "H" bus after EDG #1 re-energizes the bus? (assume NO operator action)
A. (1) 1-VS-F-58A (Filtered Exhaust Fan), THEN (2) "E" group pressurizer heaters, THEN (3) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump)
B. (1) 1-VS-F-58A (Filtered Exhaust Fan), THEN (2) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump), THEN (3) "E" group pressurizer heaters C. (1) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump), THEN (2) 1-VS-F-58A (Filtered Exhaust Fan), THEN (3) "E" group pressurizer heaters D. (1) 1-FW-P-3A (Motor Driven Auxiliary Feedwater Pump), THEN (2) "E" group pressurizer heaters, THEN (3) 1-VS-F-58A (Filtered Exhaust Fan)
K/A Emergency Diesel Generators (EDG)
Knowledge of the EDG system design feature(s) and/or interlock(s) which provide for the following: Automatic load sequencer - blackout.
(CFR: 41.7) (RO - 3.5)
K/A Match Analysis Given an operationally valid situation, the RO applicant will demonstrate knowledge of the correct functioning of the #1 EDG automatic load sequencer.
Answer Choice Analysis No concern - LP references Unit 1 components, technically 1-VS-F-58B is a unit 2 component, hence the apparent contradiction.
NOTE TO SURRY: Please carefully validate this question because your lesson plan is not completely clear about the filtered exhaust fan sequencer.
Specifically, the text implies that F-58A is the only fan that sequences on, but H/T-7.5 contradicts this.
A. INCORRECT. 1-FW-P-3A also starts 10 seconds after a loss of 2/2 RSS buses for the affected unit with a SI signal in-service. Plausible if applicant only focuses on the loss of voltage to the "H" bus.
B. INCORRECT. The MDAFW pump is the first load that is re-sequenced on to the bus for this condition. Plausible because the MDAFW pumps will sequence on to the bus at 140 seconds (between 1-VS-F-58A and "E" pressurizer heaters) if the station blackout occurs with a Hi-Hi CLS signal present.
C. CORRECT. The MDAFW pump 1-FW-P-3A starts 10 seconds after a loss of 2/2 RSS buses for the affected unit with a SI signal in-service. 1-VS-F-58A sequences onto bus "H" at 30 seconds, and the "E" group of pressurizer heaters sequences onto the bus at 180 seconds.
D. INCORRECT. See above analyses. Plausible if the applicant believes Filtered Exhaust Fans are unaffected (e.g. load sequencing is generated on the fan's alternate power source only) and would not sequence.
Supporting References
- Question is modified from a 064K4.10 question in the 2006-301 Surry ILO written exam.
- Surry lesson plan ND-90.3-LP-7, "STATION SERVICE AND EMERGENCY DISTRIBUTION PROTECTION AND CONTROL," rev. 21, p. 34-40 and section H/T-7.5.
References Provided to Applicant none
Answer: C
- 51. 0073 K5.02 1 Unit 1 Initial Conditions:
Radiography operations are in progress on a section of main steam piping.
The radiographers want to verify the correct position of the camera by using a main steam line radiation monitor located on the same elevation and close to the area where the radiography needs to take place.
To obtain a baseline reading, the camera source was placed 3.21 feet away from the radiation monitor detector. The radiation monitor read 5.92 R/hr.
Current conditions:
The camera has been moved into position to image the piping section.
Engineering calculations show that the camera should be placed 17.46 feet away from the radiation monitor detector.
The distances listed above include the difference in height from the camera to the radiation monitor detector. Consider the radiography camera as a radiation point source. Carry all calculations to three (3) decimal places.
Based on the current conditions, which ONE of the following correctly identifies the expected reading on the radiation monitor, if the camera was positioned correctly?
A. 0.037 R/hr B. 0.200 R/hr C. 1.088 R/hr D. 2.538 R/hr K/A Process Radiation Monitoring (PRM) System Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: Radiation intensity changes with source distance.
K/A Match Analysis
Given an operationally valid situation, the RO applicant is required to demonstrate knowledge of how PRM readings would change with a change in source distance.
Answer Choice Analysis A. INCORRECT. For the given situation, modeling the radiography camera as a point source, the reading will change as a ratio of the initial and final distances to the second power. This distractor is incorrect, but plausible, if the ratio is taken to the third power instead of the second. i.e. (5.92)*(3.21 ft / 17.46 ft )^3 = 0.037 R/hr B. CORRECT. Ratio the initial distance divided by the final distance, then square the fraction: (5.92)*(3.21/17.46)^2 = 0.200 R/hr.
C. INCORRECT. This distractor is plausible, because it represents the ratio taken to the first power, i.e. modeling the camera as a plane source instead of a point source.
(5.92)*(3.21/17.46) = 1.088 R/hr D. INCORRECT. This distractor is the square root of the value of the distance ratio, i.e taking the ratio to the 1/2 power: (5.92)*(3.21/17.46)^0.5 = 2.538 R/hr Supporting References
-John R. Lamarsh, Introduction to Nuclear Engineering, 2nd edition, chapter 9.11.
References Provided to Applicant none Answer: B
- 52. 0076 K4.02 3 Initial conditions:
Unit One at 100% power 1-SW-P-10A (A charging pump service water pump) in AUTO 1-SW-P-10B (B charging pump service water pump) in HAND Current conditions:
A RSST was lost due to sudden pressure All equipment operated as designed Assume sufficient time has elapsed to allow for all automatic actions to occur.
Which ONE of the following states (1) the current status of the charging pump service water pumps and (2) why?
A. (1) Both pumps currently in service (2) 1-SW-P-10B remained on the bus and 1-SW-P-10A started on low discharge header pressure.
B. (1) Both pumps currently in service (2) 1-SW-P-10B remained on the bus and 1-SW-P-10A started due to opposite emergency bus undervoltage.
C. (1) Only 1-SW-P-10B in service.
(2) 1-SW-P-10B remained on the bus and 1-SW-P-10A started on low discharge header pressure but secured upon restoration of 1-SW-P-10B D. (1) Only 1-SW-P-10A in service.
(2) 1-SW-P-10B tripped on undervoltage and 1-SW-P-10A started due to opposite emergency bus undervoltage..
K/A Service Water Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: Automatic start features associated with SWS pump controls.
K/A Match Analysis Requires knowledge of Service Water pump auto-start features.
Answer Choice Analysis A. Correct B. Incorrect: - 1st part is correct
- 2nd part is incorrect but plausible since opposite bus UV is an autostart for the charging pumps.
C. Incorrect:
- Both parts are incorrect but plausible if candidate believes that upon restoration of lead pump, the AUTO pump will secure.
D. Incorrect:
- 1st part is incorrect but plausible because numerous pumps trip on UV on associated power supply. 2nd part is supports this logic.
Supporting References ND-89.5-LP-2 Obj: B AP/12 SERVICE WATER SYSTEM ABNORMAL CONDITIONS References Provided to Applicant
none Answer: A
- 53. 01 K2.05 2 Which ONE of the following states the power supplies DIRECTLY to both Unit 1 Rod Drive MG sets and what signal would DIRECTLY open the supply breakers to the MG Sets?
A. 'A' and 'C' 260 Volt Station Service Reactor Trip B. 'A' and 'C' 480 Volt Station Service AMSAC C. 'A' and 'C' 480 Volt Station Service Reactor Trip D. 'A' and 'C' 260 Volt Station Service AMSAC K/A 001 Control Rod Drive Knowledge of bus power supplies to the following: M/G sets.
K/A Match Analysis Requires applicant to know which bus supplies Rod Drive MG set and what causes them to open.
Answer Choice Analysis A. In-Correct - Part 1 is incorrect, but plausible if the candidate confuses MG input and output power. Part 2 is incorrect, but plausible as the ultimate outcome of the MG sets losing power is a reactor trip, but a reactor trip signal will not cause AMSAC to actuate. It should be noted that AMSAC does actuate on every reactor trip from normal operating conditions.
B. Correct.
C. In-Correct - Part 1 correct. Part 2 is incorrect, but plausible as the ultimate outcome of the MG sets losing power is a reactor trip, but a reactor trip signal will not cause AMSAC to actuate. It should be noted that AMSAC does actuate on every reactor trip from normal operating conditions.
D. In-Correct - Part 1 is incorrect, but plausible if the candidate confuses MG input and output power. Part 2 is correct.
Supporting References ND-93.3-LP-3, Rod Control System, Rev. 17, Obj. D References Provided to Applicant none Answer: B
- 54. 0103 A1.01 1 A loss of coolant accident (LOCA), coincident with a failure of ALL containment spray pumps to start, causes containment pressure to INCREASE.
Which ONE of the following correctly describes the expected equipment status as containment pressure continues to rise?
A. All containment recirculation fans will operate at pressures up to 17.7 psia. At 17.7 psia, containment recirculation fans 1A and 1B will automatically trip. Containment recirculation fan 1C will continue to run at pressures greater than 17.7 psia.
B. All containment recirculation fans will operate at pressures up to 23.0 psia. At 23.0 psia, containment recirculation fans 1A and 1B will automatically trip. Containment recirculation fan 1C will continue to run at pressures greater than 23.0 psia.
C. All containment recirculation fans will operate at pressures up to 17.7 psia. At 17.7 psia, all containment recirculation fans will automatically trip.
D. All containment recirculation fans will operate at pressures up to 23.0 psia. At 23.0 psia, all containment recirculation fans will automatically trip.
K/A Containment System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls
including: Containment pressure, temperature, and humidity.
(CFR: 41.5 / 45.5) (RO - 3.7)
K/A Match Analysis The question requires the RO applicant to demonstrate recognition of expected plant conditions given an operationally valid scenario and a changing containment pressure situation.
Answer Choice Analysis A. INCORRECT. Surry lesson plan ND-88.4-LP-6 states: "the air recirculation fans will operate in containment pressure up to 8.3 psig (23 psia). At this point the 1A and 1B fans are tripped by a Hi-Hi CLS signal. This is to protect the emergency buses from an overload condition caused by the starting and running of the pumps in the SI and CLS systems. The 1C fan will remain running, since it is not powered from an emergency bus." The distractor of 17.7 psia is plausible, because it corresponds to the 3.0 psig SI actuation signal.
B. CORRECT. See above analysis.
C. INCORRECT. See above analysis. Plausible if the candidate assumes that all fans will trip to protect the fans.
D. INCORRECT. see above analyses.
Supporting References
-Modified from Surry 2002-301 test question 022A4.05.
-Surry lesson plan ND-88.4-LP-6, "CONTAINMENT VENTILATION," rev. 9, p. 5.
References Provided to Applicant none Answer: B
- 55. 0103 K4.06 1 Unit 1 Initial Conditions:
100% Power.
Current conditions:
Containment pressure transmitter 1-LM-PT-100B failed a calibration surveillance four (4) days ago.
All Technical Specification 3.7, "Instrumentation Systems," required actions for 1-LM-PT-100B have been completed.
Based on the current conditions, which ONE of the following identifies (1) the MINIMUM containment pressure, AND (2) the MINIMUM number of OPERABLE containment pressure channels that must actuate in order to close 1-RM-TV-100A/B/C (Containment Particulate and Gas Radiation Monitor 1-RM-RI-159/160 trip valves)?
A. (1) 23.0 psia (2) two (2)
B. (1) 23.0 psia (2) three (3)
C. (1) 17.7 psia (2) two (2)
D. (1) 17.7 psia (2) three (3)
K/A 103 Containment System Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following: Containment isolation system.
(CFR: 41.7) (RO - 3.1)
K/A Match Analysis Given an operationally valid scenario, the RO applicant will demonstrate knowledge of containment system design features as they relate to the containment isolation system.
Answer Choice Analysis A. INCORRECT. Correct logic, but incorrect setpoint. Plausible is applicant believes a Hi-Hi CLS is required to close the RM trip valves.
B. INCORRECT. Incorrect logic and setpoint. See analysis for C. below.
C. CORRECT. A Hi-CLS signal closes 1-TV-RM-100A/B/C. The setpoint for the Hi-CLS signal is 17.7 psia (3.0 psig). The logic for a Hi-CLS signal (with all containment pressure channels normal) is 3/4. TS 3.7 requires tripping an inoperable containment pressure channel within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (3 days), resulting in a 2/3 actuation logic for the remaining (functioning normally) channels.
D. INCORRECT. Correct pressure setpoint, but only 2/3 channels are required to actuate. Plausible if applicant does not understand CLS logic or fails to recognize TS 3.7 requires tripping (e.g. not bypassing) the inoperable channel.
Supporting References
-Surry 2006-301 exam question 103K4.06 with answer choices scrambled.
-Surry lesson plan ND-91-LP-5, "Containment Spray System," rev. 16, p. 5 and
- p. 8.
References Provided to Applicant none Answer: C
- 56. 011 K6.06 2 Unit 1 Initial Conditions:
100% Power.
Current conditions:
A controller failure causes the Unit 1 Operator to place the Charging Flow Controller to MANUAL.
The Unit 1 Operator attempts to reduce charging flow to 20 gpm to mitigate a high Pressurizer Level.
Based on the current conditions, which ONE of the following correctly describes the behavior of 1-CH-FCV-1122 when the Operator attempts to reduce charging flow to 20 gpm?
A. The Flow Limit Summator no longer limits flow, and 1-CH-FCV-1122 can be manually closed to allow 20 gpm flow.
B. The Flow Limit Summator no longer limits flow; however, 1-CH-FCV-1122 can only be manually closed to allow 25 gpm flow.
C. The Flow Limit Summator will limit charging flow to a minimum of 25 gpm.
D. The Flow Limit Summator will limit charging flow to a minimum of 30 gpm.
K/A
Pressurizer Level Control System (PZR LCS)
Knowledge of the effect of a loss or malfunction on the following will have on the PZR LCS: correlation of demand signal indication on charging pump flow valve controller to the valve position.
(CFR: 41.7 / 45.7) (RO - 2.5*)
K/A Match Analysis Given an operationally valid scenario, the RO applicant will demonstrate knowledge of the relationship between the controller output signal and the valve position with the controller in MANUAL.
Answer Choice Analysis A. CORRECT, because when the charging flow controller is in MANUAL, the Flow Limit Summator no longer limits the minimum and maximum values of charging. Therefore, FCV-1122 can be closed manually to any value.
B. INCORRECT, because when the charging flow controller is in MANUAL, the flow limit summator no longer limits the minimum and maximum values of charging. Distractor is incorrect because FCV-1122 may be manually closed to any value, even below 25 gpm flow. Distractor is plausible because the candidate may not know that FCV-1122 may be throttled to any value with the controller in MANUAL.
C. INCORRECT, because when the charging flow controller is in MANUAL, the flow limit summator no longer limits the minimum and maximum values of charging. The distractor states that the flow limit summator will limit flow, which is contrary to the fact that it will not limit flow. Distractor is plausible because candidate may not know that the flow limit summator does not function with the controller in MANUAL.
D. INCORRECT, because when the charging flow controller is in MANUAL, the flow limit summator no longer limits the maximum and minimum values of charging. The distractor states that the flow limit summator will limit flow, which is contrary to the fact that it will not limit flow. Distractor is plausible because candidate may not know that the flow limit summator does not function with the controller in MANUAL.
Supporting References
-this question is taken directly from 011K6.06 from the 2004-301 Surry exam.
-Surry lesson plan ND-93.3-LP-7, "PRESSURIZER LEVEL CONTROL SYSTEM," rev. 9, p. 6.
References Provided to Applicant none Answer: A
- 57. 014A4.01 2 Initial Unit 1 Conditions:
- Unit 1 is at 100% power
- All control rods are fully withdrawn Current Unit 1 Conditions:
- A control rod in the 'D' Control Bank drops to the bottom of the core
- Unit 1 is at 70% power
- Delta flux is within band Which ONE of the following correctly states (1) the control rod select switch position to recover the rod in accordance with 0-AP-1.01, Control Rod Misalignment AND (2) when the step counters are required to be reset in accordance with 0-AP-1.01?
A. (1) Place the ROD CONT MODE SEL SWITCH to MANUAL for rod recovery.
(2) Step counters are required to be reset prior to rod recovery.
B. (1) Place the ROD CONT MODE SEL SWITCH to the affected bank for rod recovery.
(2) Step counters are required to be reset prior to rod recovery.
C. (1) Place the ROD CONT MODE SEL SWITCH to MANUAL for rod recovery.
(2) Step counters are NOT required to be reset until after the rod is recovered.
D. (1) Place the ROD CONT MODE SEL SWITCH to the affected bank for rod recovery.
(2) Step counters are NOT required to be reset until after the rod is recovered.
K/A:
014A4.01 Rod Position Indication Ability to manually operate and/or monitor in the control room: Rod Selection Control K/A MATCH ANALYSIS:
The question requires knowledge of the rod select switch position to recover a dropped rod and knowledge of when the position indication (group counters) are required to be manipulated in conjunction with the rod control recovery.
ANSWER CHOICE ANALYSIS:
A. Incorrect. Part 1 is not correct because 0-AP-1.01 requires the mode select switch to be selected to the affected bank position. Plausible because 0-AP-1.00 requires the switch to be placed in MANUAL prior to entering 0-AP-1.01. Also plausible because the lead bank rod motion would be possible in MANUAL. Part 2 is correct.
B. Correct. The ROD CONT MODE SEL SWITCH is required to be selected to the B. affected bank (0-AP-1.01 Page 5 of 9). Also step counters are required to be reset A prior to rod recovery (0-AP-1.01 Page 7 of 9).
C Incorrect. Part 2 is incorrect because the step counters must be reset prior to rod
. recovery. Plausible because the applicant could have a misconception that the rod C should first be withdrawn prior to adjusting the step counters to read the same as
. the group.
B D Incorrect. See above.
D C
REFERENCES:
0-AP-1.01, Control Rod Misalignment, Revision 17.
1G-A6, ROD CONT SYS URGENT FAILURE, Revision 001.
ND-93.3-LP-3, Rod Control System (Handouts)
'D' Bank vice lead bank 1G-A6 would not be in, if recovering rod Answer: B
- 58. 015K5.10 3 Unit 1 plant conditions:
Unit Startup in progress following a reactor trip at Middle of Life Reactor power = 90%
1-GOP-1.5 UNIT STARTUP, 2% REACTOR POWER TO MAX ALLOWABLE POWER in progress Axial Flux Difference = 0 All Control Rods are Fully Withdrawn at 225 steps Based on the above conditions: (1) which ONE of the following states the maximum rate at which power can be increased to 100% IAW 1-GOP-1.5 (2) how will Axial Flux Difference change as power is increased ?
A. (1) 3% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) become positive
B. (1) 3% in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2) become negative C. (1) No rate limitation exists (2) become positive D. (1) No rate limitation exists (2) become negative K/A Nuclear Indication Knowledge of the operational implications of the following concepts as they apply to the NIS: Ex-core detector operation.
K/A Match Analysis Requires knowledge of Ex-Core detectors and how they are affected by plant operation.
Answer Choice Analysis A. Incorrect: 1st part is incorrect, but plausibe as this is the limitation imposed following a refueling outage. 2nd part is correct.
B. InCorrect: 1st part is incorrect, but plausibe as this is the limitation imposed following a refueling outage. As coolant passes up through the reactor core, it heats up. The hotter water causes less moderation of neutrons so power and neutron flux shifts towards the bottom of the core which causes the Axial Flux Difference to become negative.
C. Incorrect: Part 1 is correct during outage. 2nd part is plausible because for a "rodded" startup, it would be correct.
D. Correct.
Supporting References 1-GOP 1.5 Attachment 1 ND-93.2-LP-4 Obj: H References Provided to Applicant none
Answer: D
- 59. 017K5.02 2 Initial Unit 2 conditions:
- Reactor power = 100% steady state Current Unit 2 conditions:
- A LOCA is in progress E-1, Loss of Reactor or Secondary Coolant, is being performed
- All RCPs have been stopped
- Containment pressure = 47 psia and slowly increasing
- Total AFW flow = 485 gpm
- SG WR levels are: "A" = 48%, "B" = 40%, "C" = 39%
- RCS pressure = 920 psig
- IR NIs = 2E-11 amps, with SUR = 0
- CETCs indicate 600oF
- RVLIS full range = 40%
Which ONE of the following correctly states the procedure to which the control room crew is required to transition?
A. FR-C.1, Response to Inadequate Core Cooling B. FR-C.2, Response to Degraded Core Cooling C. FR-Z.1, Response to Containment High Pressure D. FR-H.5, Response to Steam Generator Low Level K/A:
017K5.02 Incore Temperature Monitoring Knowledge of the operational implications of the following concepts as they apply to the ITM system: Saturation and subcooling of water.
K/A MATCH ANALYSIS:
To arrive at the correct answer, the applicant must recognize that CET values place require a transition to FR-C.2. This procedure transition is based on incore temperature limitations which are related to subcooling conditions in the core.
ANSWER CHOICE ANALYSIS:
A. Incorrect. C.1 would only be entered if CETs = 700F. Plausible because all conditions, except for CET values would lead the applicant to C.1. Also CET values are well above normal which would indicate an issue with core cooling.
B. Correct. Orange Path on Core Cooling. CET < 1200F / SCM < 85F / No RCPs on / CET < 700 and RVLIS < 46%.
C. Incorrect. Conditions for Z.1 on an Orange Path are met; however, C.2 Orange Path is a higher priority. Plausible because the conditions are met for Z.1.
D. Incorrect. Conditions are met for H.5 on a Yellow Path; however, C.2 is required to be entered first. Plausible because the conditions are met for H.5.
Surry Exam Bank Question #1066 Previous Surry Exam 2002-301 (Not one of the previous 2 exams)
REFERENCES:
Surry Safety Functions Status Trees Answer: B
- 60. 035A2.01 2 Unit 1 initial conditions:
Reactor trip from 40% power SI actuated 1-E-0 REACTOR TRIP OR SAFETY INJECTION initiated Current plant conditions:
A NR SG level = 26% decreasing B NR SG level = 22% decreasing C NR SG level = 29% decreasing Based on the above plant conditions, which ONE of the following correctly states (1) the MINIMUM SG level at which the first signal to start an AFW pump occurs (assuming no operator action) and (2) the MINIMUM required Steam Generator Narrow Range Level that allows SI flow reduction IAW 1-E-0 REACTOR TRIP OR SAFETY INJECTION?
A. (1) 13%
(2) Greater than 12%
B. (1) 13%
(2) Greater than 22%
C. (1) 17%
(2) Greater than 12%
D. (1) 17%
(2) Greater than 22%
K/A Steam Generator Ability to (a) predict the impacts of the following malfunctions or operations on the S/Gs; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Reactor trip /
K/A Match Analysis Requires knowledge of how a reactor trip w/SI affects SG levels and how level affects procedure use.
Answer Choice Analysis A. Incorrect: 1st part is plausible because 13% is the AMSAC setpt to start AFW pumps. 2nd part is correct.
B. Incorrect: 1st part is plausible because 13% is the AMSAC setpt to start AFW pumps. 2nd part is plausible because 22% is the bottom of the control band (22-50%) used in 1-E-0 if SI flow is not throttled.
C. Correct: AFW pumps receive a start signal on LO-LO level of 17% in any 2/3 Sgs.
12% level in any SG or total feed flow > 350 gpm is used as part of the SI throttle criteria.
D. Incorrect: 1st part is correct. 2nd part is plausible because 22% is the bottom of the control band (22-50%) used in 1-E-0 if SI flow is not throttled.
Supporting References ND-89.3-LP-4 Aux FDW ND-95.3-LP-3 E-0 Obj C References Provided to Applicant none Answer: C
- 61. 041 A3.03 2 Unit 1 Initial Conditions:
100% power.
Rod control is selected to P-446, Channel III turbine first stage impulse
pressure.
P-447, Channel IV turbine first stage impulse pressure, fails LOW.
Annunciator 1H-D7, "STM DUMP PERM," is lit.
Current conditions:
No operator actions have been performed to address the P-447 failure.
Control rods are INSERTING in automatic.
Steam flow on all channels is INCREASING.
Based on the current conditions, which ONE of the following correctly identifies the cause?
A. A main steam line safety valve has lifted and will not reseat.
B. Median Tave has failed HIGH.
C. P-446 has failed HIGH.
D. P-464, Steam header pressure, has failed HIGH.
K/A Steam Dump/Turbine Bypass Control Ability to monitor automatic operation of the SDS, including: Steam flow.
(CFR: 41.7 / 45.5) (RO - 2.7)
K/A Match Analysis The question gives the RO applicant an opportunity to demonstrate integrated knowledge of the steam dump control system, given a set of operational conditions including abnormal steam flow.
Answer Choice Analysis A. INCORRECT. The steam line safety vale may cause the increased steam flow indication, but the increase in steam demand would cause Tave to DECREASE, which would tend to cause automatic rod control to move rods OUT (rather than in).
B. CORRECT. Tave would indicate high, and with the steam dumps armed, the dumps would open, causing steam flow to increase. The false high Tave would also cause rods to drive in.
C. INCORRECT. P-446 failing LOW would cause steam dumps to open;
however, failing HIGH will keep steam dumps from opening (plausible but incorrect).
D. INCORRECT. P-464 failing HIGH would cause everything listed to happen, if the steam dumps were in pressure control mode, but they are still in the Tave control mode.
Supporting References
-Surry lesson plan ND-93.3-LP-9. "STEAM DUMPS," rev. 13, p. 6,7,21.
-Surry lesson plan ND-93.3-LP-2, "DELTA T/TAVG INSTRUMENTATION SYSTEM," rev. 10, p. handouts/diagrams pages.
References Provided to Applicant none Answer: B
- 62. 055 G2.4.45 2 Unit 1 Initial Conditions:
Unit was at 28% power when condenser vacuum began to degrade.
Current conditions:
Annunciator 1E-E3, "DELTA FLUX DEVIATION," is lit.
Annunciator 1G-H8, "ROD BANK D EXTRA LO LIMIT," is lit.
Annunciator 1F-B6, "TURB LO VAC," has been lit for five (5) minutes.
Control rods are inserting in automatic.
Condenser vacuum continues to DECREASE with no signs of recovery.
Turbine is at 14% load and DECREASING.
Based on the current conditions, which ONE of the following identifies the required operator action, in accordance with 1-AP-14.00, "LOSS OF MAIN CONDENSER VACUUM?"
A. Commence an emergency boration using 1-AP-3.00, "EMERGENCY BORATION."
B. IMMEDIATELY trip the Reactor and enter 1-E-0, "REACTOR TRIP OR SAFETY INJECTION."
C. IMMEDIATELY trip the Turbine and stabilize the unit using the steam dumps.
D. IF condenser vacuum is less than 24.5 in-Hg for a five (5) minute period, THEN trip the Turbine and stabilize the unit using the steam dumps.
K/A Condenser Air Removal Ability to prioritize and interpret the significance of each annunciator or alarm.
(CFR: 41.10 / 43.5 / 45.3 / 45.12) (RO - 4.1)
K/A Match Analysis Based on an operationally valid set of plant conditions and alarms associated with condenser air removal, recognize entry conditions into E-0.
Answer Choice Analysis A. INCORRECT. Must trip the reactor and GO TO 1-E-0 based on being less than 30% turbine power with condenser vacuum less than 26.5 in Hg for 5 minutes with no signs of recovery. This is recognized by the TURB LO VAC alarm being "in" for 5 minutes (setpoint is 25 in Hg vacuum). Plausible because of the LO-LO insertion limit alarm in the question stem. Also technically incorrect because the emergency boration is directed in 1-AP-14.00, and NOT in 1-AP-3.00.
B. CORRECT. Must trip the reactor and GO TO 1-E-0 based on being less than 30% turbine power with condenser vacuum less than 26.5 in Hg for 5 minutes with no signs of recovery. This is recognized by the TURB LO VAC alarm being "in" for 5 minutes (setpoint is 25 in Hg vacuum).
C. INCORRECT. A reactor trip is required based on condenser vacuum as described above. This statement is plausible, because it would be correct if turbine power/reactor power were less than 10% (or if the output breakers were open).
D. INCORRECT. See above analysis. Distractor is plausible, because it is the correct reasoning when power is greater than 30%; however, must trip the Rx--
not the turbine. I want this distractor to basically improve the plausibility of C.
Supporting References AP-14.00, "LOSS OF MAIN CONDENSER VACUUM," rev. 5, esp. attachment 3.
-Surry lesson plan ND-95.1-LP-6, "LOSS OF CONDENSER VACUUM," rev. 10.
References Provided to Applicant none Answer: B
- 63. 072 K4.01 1 Unit 1 initial conditions:
Shut down for refueling Containment purge in progress Current plant conditions:
A High radiation signal on the containment particulate (1-RM-RI-159 CTMT PARTC) radiation monitor occurs Based on the above conditions which ONE of the following correctly states the status of the containment purge components?
A. Containment purge supply fans (1-VS-F-4A and B) off Containment purge supply MOVs (1-VS-MOV-100A and B) closed, Containment purge discharge MOVs (1-VS-MOV-100C and D) closed B. Containment purge supply fans (1-VS-F-4A and B) off Containment purge supply MOVs (1-VS-MOV-100A and B) closed, Containment purge discharge MOVs (1-VS-MOV-100C and D) open C. Containment purge supply fans (1-VS-F-4A and B) on Containment purge supply MOVs (1-VS-MOV-100A and B) closed, Containment purge discharge MOVs (1-VS-MOV-100C and D) open D. Containment purge supply fans (1-VS-F-4A and B) on Containment purge supply MOVs (1-VS-MOV-100A and B) open, Containment purge discharge MOVs (1-VS-MOV-100C and D) open K/A Area Radiation Monitoring.
Knowledge of ARM system design feature(s) and/or interlock(s) which provide for the following: containment ventilation isolation.
K/A Match Analysis Requires knowledge of ARM system interactions/interlocks with the containment purge system.
Answer Choice Analysis
A. Correct. Upon receiving the radiation alarm, the purge supply fan trips and the containment isolation MOVs close to prevent any release to the atmosphere.
B. Incorrect: Discharge dampers will be closed. Plausible because these dampers remaining open will allow the escaping air to be filtered and monitored.
C. Incorrect: Discharge dampers will be closed. Plausible because these dampers remaining open will allow the escaping air to be filtered and monitored.
D. Incorrect: All of the stated dampers will be closed and the purge fan off.
Plausible because if a different containment ARM alarmed, it would not control this function and it would be correct.
Supporting References ND-88.4-LP-6 Obj D References Provided to Applicant none Licensee verify Containment Purge is not a subsystem of Containment Vent.
Answer: A
- 64. 078 K3.03 2 Plant initial conditions:
Reactor Power = 100% both units Instrument Air Systems are in their NORMAL alignments (split out)
Operator reports an air leak on Unit 1 instrument air header Unit 1 instrument air pressure is currently 85 psig and DECREASING Based on the above conditions, which ONE of the following correctly states the expected status of (1) Unit 2 Instrument Air pressure and (2) the Unit 1 Instrument Air Compressor?
A. (1) Unit 2 Instrument Air pressure will decrease until air pressure equals 80 psig.
(2) operating.
B. (1) Unit 2 Instrument Air pressure will decrease until air pressure equals 80 psig.
(2) NOT operating.
C. (1) Unit 2 Instrument Air pressure will decrease until 2-IA-C-1 starts.
(2) operating.
D. (1) Unit 2 Instrument Air pressure will decrease until 2-IA-C-1 starts.
(2) NOT operating.
K/A Instrument Air Knowledge of the effect that a loss of malfunction of the IAS will have on the following: Cross-tied units.
K/A Match Analysis Requires knowledge of lineups of the IA system and how a leak on one will affect the other.
Answer Choice Analysis A. Incorrect: Unit IA headers are normally split so a loss of IA on one unit does not affect the other. Plausible because Service Air headers are normally cross connected. In AUTO, the IA compressor will start if IA header pressure decreases to 90 psig. Annunciator setpt is 80 psig.
B. Incorrect: Unit IA headers are normally split so a loss of IA on one unit does not affect the other. Plausible because Service Air headers are normally cross connected. 2nd part is plausible because there numerous setpts for the IA system both above and below 90 psig.
C. Correct: Instrument air headers are normally split 2nd part is correct.
D. Incorrect: 1st part is plausible because IA is kept split from containment air system.
2nd part is plausible because there numerous setpts for the IA system both above and below 90 psig.
Supporting References ND-92.1 LP1 Station Air Systems Obj: B, E. Annunciator Response 1B-E6 IA LO HDS PRESS/IA COMPR 1 TRBL References Provided to Applicant none utility to provide instances when the unit air headers may be split - just state they are split.
Answer: C
- 65. 086 A1.05 2 Which ONE of the following states (1) the capacity of each Fire Water Tank and (2) if a
domestic water leak occurred, the tank level at which the leak would stop?
A. (1) 250,000 gallons per tank (2) 200,000 gallons per tank B. (1) 300,000 gallons per tank (2) 250,000 gallons per tank C. (1) 300,000 gallons per tank (2) 200,000 gallons per tank D. (1) 250,000 gallons per tank (2) 50,000 gallons per tank K/A Fire Protection System (FPS)
Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Fire Protection System operating the controls including:
Fire Water Storage Tank Level K/A Match Analysis Requires knowledge of tank capcity and amount required for fire protection Answer Choice Analysis A. INCORRECT. Part 1 Incorrect, but plausible if the candidate confuses tank capacity with amount of water required for fire protection purposes. Part 2 is plausible if the candidate recalls that 50,000 gpm per tank is for domestic water.
B. CORRECT. See attached calculation sheet.
C. INCORRECT. Part 1 is correct. Part 2 is plausible if the candidate doubles the amount of water available for domestic water.
D. INCORRECT. Part 1 is correct. Part 2 is plausible if the candidate believes the tap is located at the bottom of the tank vice top.
Supporting References
- Surry TRM TR 3.7.1, "Fire Suppression Water System," rev. 19, p. 3.7.1-1
- Surry lesson plan ND-92.2-LP-1, "FIRE PROTECTION SYSTEMS," rev. 12, especially p. 6.
References Provided to Applicant none Answer: B
- 66. G2.1.29 1 Which ONE of the following (1) correctly states the maximum allowable length of a valve wrench used IAW OP-AA-100, Conduct of Operations, AND (2) whether OP-AA-100 allows a valve wrench to be used on manual valves as well as motor operated valves (MOVs)?
A. (1) Valve wrench length is limited to approximately 1.5 times the handwheel diameter.
(2) Valve wrench is permitted to be used on manual valves but not MOVs.
B. (1) Valve wrench length is limited to approximately 2.0 times the handwheel diameter.
(2) Valve wrench is permitted to be used on manual valves but not MOVs.
C. (1) Valve wrench length is limited to approximately 1.5 times the handwheel diameter.
(2) Valve wrench is permitted to be used on both manual valves and MOVs.
D. (1) Valve wrench length is limited to approximately 2.0 times the handwheel diameter.
(2) Valve wrench is permitted to be used on both manual valves and MOVs.
K/A Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.
K/A Match Analysis Requires applicant to know the restrictions on use of additional force when aligning valves.
Answer Choice Analysis A. Correct - A lever up to 1 1/2 times the handwheel diameter may be used to open manual valves, but is not to be used on MOVs.
B. In-Correct but plausible since leverage is allowed, but the restriction limits the size to 1 1/2 times. 2nd half of response is correct.
C. In-Correct but plausible since a lever up to 1 1/2 times the handwheel diameter may be used to open the valve. However, use of leverage is only allowed on manual valves.
D. In-Correct but plausible since leverage is allowed to open the valve, but the lever is limited to 1 1/2 times the diameter of the handwheel. In addition, use of leverage is only allowed on manual valves.
Supporting References OP-AA-100, Conduct of Operations. Rev. 05 References Provided to Applicant none Answer: A
- 67. G2.1.40 1 Unit 1 Initial Conditions:
Core re-fueling operations are in progress.
Approximately 3/4 of the new core has been loaded without incident.
Current conditions:
One Source Range count rate is double (2X) the initial reference value.
The other Source Range count rate is (1.75X) (less than double) the initial reference value.
The 1/M plot is approaching 0.65.
Based on the current conditions, which ONE of the following identifies the MINIMUM conditions that would require stopping core alterations, in accordance with the Precautions and Limitations of 1-OP-FH-001, "CONTROLLING PROCEDURE FOR REFUELING?"
A. Core alterations are required to be stopped immediately and subcriticality reevaluated.
B. Core alterations may continue, but IF BOTH Source Range count rates reach one doubling from the reference value, then core alterations are required to be stopped immediately and subcriticality reevaluated.
C. Core alterations may continue, but IF the 1/M plot approaches 0.5, then core alterations are required to be stopped immediately and subcriticality reevaluated.
D. Core alterations may continue, but IF BOTH Source Range count rates reach one doubling from the reference value, AND the 1/M plot approaches 0.5, then core alterations are required to be stopped immediately and subcriticality reevaluated.
K/A Generic topic K/A:
Knowledge of refueling administrative requirements.
(CFR: 41.10 / 43.5 / 45.13) (RO - 2.8)
K/A Match Analysis The question requires the RO applicant to demonstrate knowledge of an important precaution and limitation with respect to reactivity control during refueling operations.
Answer Choice Analysis A. CORRECT. Precaution and Limitation 4.49 of procedure 1-OP-FH-001, "CONTROLLING PROCEDURE FOR REFUELING," states: "If the Source Range count rate on either detector doubles from the reference value, or the 1/M plot approaches 0.5, all core alterations must be stopped immediately and subcriticality reevaluated." The additional distractors are incorrect, but plausible, combinations of the sections of this P&L.
B. INCORRECT. See analysis of A. above.
C. INCORRECT. See analysis of A. above.
D. INCORRECT. See analysis of A. above.
Supporting References
-Surry procedure 1-OP-FH-001, "CONTROLLING PROCEDURE FOR REFUELING," rev. 21, p. 18 of 103, step 4.49.
References Provided to Applicant none Answer: A
- 68. G2.1.44 2 Unit 1 initial conditions:
Core re-load in progress SR NI background count rate = 10 cps Current plant conditions:
1G-C1, NIS SOURCE RNG SHUTDN HI FLUX, alarms Based on the above conditions, which ONE of the following correctly states (1) the minimum count rate that would cause the alarm and (2) what actions are directed by ARP 1G-C1?
A. (1) 42 cps (2) Direct the refueling SRO to place fuel in a safe condition and evacuate containment.
B. (1) 42 cps (2) Emergency borate and direct the refueling SRO to stop all refueling activities.
C. (1) 60 cps (2) Direct the refueling SRO to place fuel in a safe condition and evacuate containment.
D. (1) 60 cps (2) Emergency borate and direct the refueling SRO to stop all refueling activities.
K/A Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.
K/A Match Analysis Requires knowledge of MCR indications and actions for alarms during refueling activities.
Answer Choice Analysis A. Correct: Per 1G-C1, alarm setting = 0.5 decades above background, place fuel in a safe conditions, and evacuate containment.
B. Incorrect: 1st part is correct. 2nd part is incorrect because you are not directed to emergency borate. 2nd part is plausible because if this conditinos occurred at a hot zero power, it may be correct.
C. Incorrect: 1st part is incorrect because the setting is 0.5 decades above background. 0.5 decases on a log scale = 3.16 or about 32 counts above background. 1st part is plausible because 0.5 decades on a linear scale is 50 counts above background. 2nd part is correct.
D. Incorrect: 1st part is incorrect because the setting is 0.5 decades above background. 0.5 decases on a log scale = 3.16 or about 32 counts above background. 1st part is plausible because 0.5 decades on a linear scale is 50 counts above background. 2nd part is incorrect because you are not directed to emergency borate. 2nd part is plausible because if this conditinos occurred at a hot zero power, it may be correct.
Supporting References 1G-C1 Alarm Response Guide NIS SOURCE RNG SHUTDN HI FLUX References Provided to Applicant none Licensee to verify that emergency boration would not be a method used to "borate as necessary" if Shutdown Margin margin was determined to be inadequate".
Answer: A
- 69. G2.2.40 3 With the unit initially at 100% power, 1-MS-PT-1446 (Ch III Impulse pressure) fails to 0%. Unit conditions are as follows:
Reactor power 95% - by delta-T D bank control rods at 212 steps Delta Flux is currently -7 Delta flux target is -1.5 Tave is 568F Tref is 572F Steam generator levels are stable at 33% narrow range.
RCS pressure is currently 2200 psig Which ONE of the following states the MOST LIMITING LCO (if any), and required actions?
A. No LCO actions are required.
B. 15 minute clock to restore delta-flux in band due to delta flux being outside target band.
C. 30 minute clock to restore pressurizer pressure as pressurizer pressure is outside the allowable band.
D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> clock to verify permissive status due to pimp failure to P-10 and P-7 interlocks K/A Tier 3: Generic K/A Ability to apply Technical Specifications for a system.ck is most limiting.
Answer Choice Analysis A. INCORRECT. No obvious TS given.
B. CORRECT.
C. INCORRECT. Plausible as this RCS pressure is below Tech Spec requirements, but the clock requirement is 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, not 30 minutes.
D. INCORRECT. Plausible as this is a valid TS for this scenario. Normally this is the most limiting TS on a Pimp failure.
Supporting References
-Surry Technical Specification 3.7, "INSTRUMENTATION SYSTEMS,"
References Provided to Applicant none Answer: B
- 70. G2.2.42 2 Unit One at 100% power and stable.
Chemistry morning report has been issued with the following parameters given:
- Unit One "A" safety injection accumulator (1-SI-TK-1A) - boron concentration- 2235 ppm.
- Unit One "B" safety injection accumulator (1-SI-TK-1B) - boron concentration- 2500 ppm.
- Unit One "C" safety injection accumulator (1-SI-TK-1C) - boron concentration- 2300 ppm.
- "A" Waste Gas Decay Tank- 1.65% oxygen concentration.
- "A" Waste Gas Decay Tank- 25,720 curies.
Which ONE of the following states ALL the above parameters that require entry into a Technical Specification LCO?
A. "A" accumulator boron and "A" Waste Gas Decay Tank O2 content B. "B" accumulator boron and "A" Waste Gas Decay Tank O2 content C. "A" accumulator boron and "A" Waste Gas Decay Tank curie content.
D. "B" accumulator boron and "A" Waste Gas Decay Tank curie content.
K/A Tier 3: Generic Ability to recognize system parameters that are entry-level conditions for Technical Specifications.
(CFR: 41.7 / 41.10 / 43.2 / 43.3 / 45.3) (RO - 3.9)
K/A Match Analysis Question requires analysis of given conditions and recognize conditions that are not in compliance with Tech Spec requirements.
Answer Choice Analysis A. INCORRECT. Part concerning "A" accumulator boron is correct, but second half is incorrect, but plausible as 1.65% oxygen is the alarm setpoint - 2% is the threshold as per tech specs.
B. INCORRECT. Part 1 is incorrect but plausible as given boron concentration is unusually high - and this boron is the high spec for the RWST boron (same page on tech specs).
C. CORRECT.
D. CORRECT. First part is incorrect (see distracter "B" analysis), second part is correct.
Supporting References
-Surry Technical Specification 3.3 and section 3.11.
-New Question References Provided to Applicant none Answer: C
- 71. G2.3.5 2 You are assigned to oversee work being performed in a Radiation area.
Which ONE of the following describes: (1) the types of radiation that are measured by the DAD and (2) the requirements for DAD placement if work is to be performed in a contaminated area?
A. (1) Gamma & X-Ray ONLY (2) Inside protective clothing with TLD B. (1) Gamma & X-Ray ONLY (2) Outside protective clothing in a whirlpack C. (1) Gamma, Beta and Neutron (2) Inside protective clothing with TLD D. (1) Gamma, Beta and Neutron (2) Outside protective clothing in a whirlpack K/A Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.
K/A Match Analysis Requires knowledge of how to use personnel monitoring equipment.
Answer Choice Analysis A. Correct. 1st part is correct. 2nd part is incorrect because the candidate may believe that the alarm is a buffer to his/her actual limit.
B. Correct C. Incorrect: 1st part is plausible because if asked about the TLD, it would be correct.
2nd part is correct.
D. Incorrect: 1st part is plausible because if asked about the TLD, it would be correct.
2nd part is incorrect because you just have to move to an area where the alarm stops. It is plausible because the dose alarm does require you to go to HP.
Supporting References ND-81.2-LP2, Obj: C References Provided to Applicant none Answer: B
- 72. G2.3.7 2 While taking LOGS in the auxiliary building, a mechanic, who is performing an overhaul on 1-CH-P-1A (A charging pump), approaches you and asks for assistance in lifting the auxiliary oil pump. He states that he will only require your assistance for 20-30 minutes.
Which ONE of the following states the proper response to this request?
A. Provide assistance and when logs are complete, ask health physics to assign the dose received while helping the mechanic to the mechanic's RWP.
B. Render the requested assistance on Operations RWP as long as the dose received will not cause you to reach either your DOSE RATE LIMIT or DOSE LIMIT.
C. Inform the mechanic that you are unable to render the requested assistance while on the current Operations RWP.
D. Call health physics shift supervisor and request to be placed on the mechanics RWP. When complete, contact the health physics shift supervisor again, and get reassigned to the normal operations RWP.
K/A Tier 3: Generic.
Ability to comply with radiation work permit requirements during normal or abnormal conditions.
(CFR: 41.12/45.10) (RO - 3.5)
K/A Match Analysis The question requires the understanding of the RWP usage requirements.
Answer Choice Analysis A- INCORRECT - a worker cannot perform a task in the charging pump cubicle without the proper briefing.
B- INCORRECT - work can only be performed in the RCA for the RWP that the worker is assigned.
C- CORRECT D- INCORRECT - DAD assignments are made to a specific RWP - this cannot be changed over the phone. Plausible from an ALARA aspect.
Supporting References References Provided to Applicant
- None Answer: C
- 73. G2.4.14 2 Unit 1 initial conditions:
Reactor Trip Critical safety functions as follows.
SUB CRITICALITY - GREEN HEAT SINK - ORANGE CORE COOLING - ORANGE INVENTO RY - YELLOW CONTAINMENT - RED INTEGRITY - ORANGE Based on the above conditions, when addressing Critical Safety Functions (CSFs) which ONE of the following CSFs has the highest priority and should therefore be addressed first?
A. Heat Sink
B. Core Cooling C. Containment D. Integrity K/A Knowledge of general guidelines for EOP usage.
K/A Match Analysis Requires knowledge of prioritizing conditions when using the EOP.
Answer Choice Analysis A. Incorrect: Plausible because it is a higher priorityCSF.
B. Incorrect: Plausible because it is a higher priority CSF. Would be correct if no Red condition existed.
C. Correct: Red gives it the higher priority.
D. Incorrect: Plausible because it is a higher priority CSF.
Supporting References ND-95.3-LP-26 Sect. B & C, Obj: B & C References Provided to Applicant none Answer: C
- 74. G2.4.18 2 Unit 1 initial conditions:
Loss of all feedwater has occurred from 100% power EOPs are progress Current plant conditions:
Transition to 1-FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK has just been made 1A Wide Range SG level = 4% decreasing 1B Wide Range SG level = 5% decreasing 1C Wide Range SG level = 6% decreasing RCS pressure = 2300 psig increasing CETC = 580 oF increasing All RCPs are secured
Based on the above conditions: (1) which ONE of the following actions are directed by 1-FR-H.1 and (2) why?
A. (1) Commence bleed and feed (2) At least Two SGs are considered dry so transition to another form of decay heat removal must be made before conditions degrade further B. (1) Commence bleed and feed (2) RCS pressure may reach pressurizer safety valve setpoints, so transition to another form of decay heat removal must be made to prevent water relief through the safety valves.
C. (1) Cross Connect with Unit 2 AFW and feed at the maximum available rate (2) To reduce RCS temperature to < 550 oF for establishing a heat sink D. (1) Cross Connect with Unit 2 AFW and feed at the maximum available rate (2) To increase SG level to > 7% in any SG for establishing a heat sink K/A Knowledge of the specific bases for EOPs.
K/A Match Analysis Requires knowledge of the specific Basis for EOP steps.
Answer Choice Analysis A. Correct. Caution in 1-FR-H.1 step 2: If WIDE RANGE level in any 2 SGs is less than 7% [22%] OR PRZR pressure is greater than or equal to 2335 psig due to loss of secondary heat sink, RCPs should be tripped and Steps 11 through 18 should be immediately initiated for bleed and feed.
B. Incorrect: 1st part is correct: 2nd part is incorrect. 2nd part is plausible because water relief thru safeties is not desired.
C. Incorrect: 1st part is incorrect because the criteria for commencing Feed and Bleed has been met. 1st part is plausible because it would be correct if 2 SG levels were > 7%. 2nd would be correct for a dry SG.
D. Incorrect: 1st part is incorrect because the criteria for commencing Feed and Bleed has been met. 1st part is plausible because it would be correct if 2 SG levels were > 7%. 2nd would be correct for a dry SG.
Supporting References EOP 1-FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK ND-95.3-LP41 Response to Loss of Secondary Heat Sink B.4 Obj: B References Provided to Applicant none Answer: A
- 75. G2.4.49 3 Which ONE of the following describes that actions required on a failure of the reactor to trip (ATWS) in accordance with 1-FR-S.1 (Response to Nuclear Power Generation/ATWS)?
A. Place rod control in MANUAL and manually trip the turbine. If turbine will not trip, then close the main steam trip valves. Manually insert control rods.
B. Place rod control in MANUAL and manually trip the turbine. If turbine will not trip, then reduce limiter to zero. Manually insert control rods.
C. Place rod control in AUTOMATIC and manually trip the turbine. If turbine will not trip, then reduce limiter to zero. Verify automatic rod insertion.
D. Place rod control in AUTOMATIC and manually trip the turbine. If turbine will not trip, then close the main steam trip valves. Verify automatic rod insertion.
K/A Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
K/A Match Analysis Requires knowledge of immediate actions.
Answer Choice Analysis A. Incorrect: plausible (rods to manual) if candidate confuses step 3 rno actions into step 1. Closing the MSTVs is plausible as it is part of the RNO for turbine not tripping, but after limiter reduction.
B. Incorrect. plausible (rods to manual) if candidate confuses step 3 rno actions into step 1.
C. Correct D. Incorrect: plausible - Closing the MSTVs is plausible as it is part of the RNO for turbine not tripping, but after limiter reduction.
Supporting References OP AA 100 Sect. 4. Immediate Actions FR-S.1 References Provided to Applicant none Answer: C
- 1. 0026G2.1.7 2 Unit 1 Initial Conditions:
The operations team is cooling down the unit in preparation for refueling in accordance with 1-GOP-2.6, "UNIT COOLDOWN, LESS THAN 205 °F TO AMBIENT."
The pressurizer (PRZR) level is at 70%.
All pressurizer heaters are available.
RCS Pressure is approximately 250 psig.
RCS Temperature is approximately 180 °F.
'A' and 'B' steam generators narrow range levels are 50%, while 'C' steam generator is at 98%
Auxiliary feedwater is available to all SGs.
All RCS temperature indicators are available.
All three steam generator PORVs are operable.
All RCPs are stopped.
Current conditions:
A large unisolable Component Cooling Water leak caused a complete and sustained loss of Component Cooling Water.
The operations team entered 1-AP-27.00, "LOSS OF DECAY HEAT REMOVAL CAPABILITY."
CETC temperatures are approaching saturation.
Based on the current conditions, which ONE of the following is the FIRST method of providing decay heat removal, in accordance with 1-AP-27.00?
A. Forced feed cooling.
B. Reflux boiling heat removal.
C. Natural Circulation.
D. Cooling the RCS with the SFP and RWST coolers.
K/A Loss of Component Cooling Water: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
(CFR: 41.5/43.5/45.12/45.13) (SRO - 4.7)
K/A Match Analysis Given a complete loss of component cooling water under S/D and C/D conditions, the applicant must use the plant conditions to determine the appropriate course of action.
SRO-Only Analysis See attached SRO-only guidance flowchart. As an amplification, this question is focusing on the correct procedural selection of the various attachments in AP-27.00 (the four answer choices are word-for-word the titles of the various attachments in AP-27.00); and is therefore testing procedural knowledge on a different and more detailed level than what is expected for a RO.
Answer Choice Analysis A. INCORRECT.
B. INCORRECT.
C. CORRECT.
D. INCORRECT.
Supporting References GOP-2.6, "UNIT COOLDOWN, LESS THAN 205 F TO AMBIENT," rev 28 (p.
8, 12, 18, 19, 20-22)
-SPS TS Fig. 3.1-2, "RCS COOLDOWN LIMITATIONS." AP-15.00, "LOSS OF COMPONENT COOLING," CAUTION before step 1.
1-AP-27.00, "LOSS OF DECAY HEAT REMOVAL CAPABILITY," rev 18; procedural flowpath to steps 19, 20, and 21; attachments 4, 5, 6 1-OSP-ZZ-004, "UNIT 1 SAFETY SYSTEMS STATUS LIST FOR COLD SHUTDOWN/REFUELING CONDITIONS," rev 35, p. 10 (table of mandatory and non-mandatory backup cooling methods)
References Provided to Applicant none This question appears to be testing the in-depth knowledge of AP-27.00 and OSP-ZZ-004 vice a loss of component cooling water.
All three SGs need to be available in accordance with AP-27.00 for reflux cooling. If natural circulation is unavailable (need a reason - flow
obstruction,etc...), how is reflux boiling available?
Note: OSP-ZZ-004 for the given plant conditions lists Natural Circ and Forced Feed and Bleed were the mandatory back-ups.
Recommend alternate question.
Answer: C
- 2. 0036AA2.03 2 Which ONE of the following correctly describes (1) the definition of the Exclusion Area Boundary for Surry Power Station, and (2) the basis for the Exclusion Area Boundary size?
A. (1) A 1650-foot radius circle centered at the Unit 1 reactor containment building (2) An individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not a total radiation dose in excess of 300 Rem to the thyroid from iodine exposure.
B. (1) A 1650-foot radius circle centered at the Unit 1 reactor containment building (2) An individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not a total radiation dose in excess of 50 Rem to the thyroid from iodine exposure.
C. (1) A 1650-foot radius circle centered at the Unit 2 reactor containment building (2) An individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not a total radiation dose in excess of 300 Rem to the thyroid from iodine exposure.
D. (1) A 1650-foot radius circle centered at the Unit 2 reactor containment building (2) An individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not a total radiation dose in excess of 50 Rem to the thyroid from iodine exposure.
K/A Ability to determine and interpret the following as they apply to the Fuel Handling Incidents: Magnitude of potential radioactive release.
(CFR: 43.5/45.13) (SRO - 4.2)
K/A Match Analysis The question requires the applicant to understand the assumptions that are
behind the fuel handling accident (FHA) analysis as presented in the basis of the SPS TS 3.10.
SRO-Only Analysis The applicant is required to know and understand the basis for actions required by technical specification during fuel handling accidents (outside the knowledge requirement for ROs).
Answer Choice Analysis A- CORRECT.
B- INCORRECT.
C- INCORRECT.
D- INCORRECT.
Supporting References
-Surry Power Station Technical Specifications 3.10 (basis).
References Provided to Applicant none Answer: A
- 3. 0039A2.03 2 Unit 1 Initial Conditions:
100% Power A tube leak in the 'B' Steam Generator (S/G) has been identified.
Control room operators have transitioned to 1-AP-24.00, "MINOR SG TUBE LEAK."
Current conditions:
Condenser air ejector radiation monitor, RI-SV-111, alarms but the automatic actions do NOT occur.
Main Steam (MS) Line B radiation monitor, RI-MS-125, alarms.
MS Line A and C radiation monitor readings are slightly higher than before.
The Senior Reactor Operator directs a manual reactor trip and initiation of 1-E-0, "REACTOR TRIP OR SAFETY INJECTION."
Safety Injection (SI) does NOT automatically actuate.
At step 4 of 1-E-0, it is determined that SI is NOT REQUIRED.
Based on the current conditions, which ONE of the following is (1) the correct procedural flowpath, AND (2) the correct method to procedurally address the failure of RV-SI-111 automatic actions?
A. (1) Transition to 1-ES-0.1(REACTOR TRIP RESPONSE) and perform 1-AP-24.00 (MINOR STEAM GENERATOR TUBE LEAK) in parallel (2) Perform steps in 1-AP-24.00 to correct the failure of RV-SI-111 automatic actions, in parallel with 1-ES-0.1.
B. (1) Transition to 1-ES-0.1(REACTOR TRIP RESPONSE) and perform 1-AP-24.01 (LARGE STEAM GENERATOR TUBE LEAK) in parallel (2) Perform steps in 1-AP-24.01 to correct the failure of RV-SI-111 automatic actions, in parallel with 1-ES-0.1.
C. (1) Transition to 1-ES-0.1, (REACTOR TRIP RESPONSE) and then transition to 1-AP-24.01 (LARGE STEAM GENERATOR TUBE LEAK).
(2) Perform steps in 1-AP-24.01 to correct the failure of RV-SI-111 automatic actions.
D. (1) Transition to 1-ES-0.1, (REACTOR TRIP RESPONSE) and then transition to 1-AP-24.01 (LARGE STEAM GENERATOR TUBE LEAK).
(2) Perform steps in 1-AP-24.00 to correct the failure of RV-SI-111 automatic actions, in parallel with 1-AP-24.01.
K/A Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Indications and alarms for main steam and area radiation monitors (during SGTR).
(CFR: 41.5/43.5/45.3/45.13) (SRO - 3.7)
K/A Match Analysis Requires the applicant to identify the situation, given a set of conditions, and exercise the correct procedures to mitigate both the SGTR and a failure of SJAE radiation monitor automatic actions.
SRO-Only Analysis
See attached SRO-only guidance flowchart. Internal EOP/AP procedure transition. Knowledge beyond simply entry conditions is required to arrive at the correct answer.
Answer Choice Analysis A. INCORRECT. Both AP-24.00 and AP-24.01 clearly state that the correct transition is to AP-24.01 instead of ES-0.1. However, ES-0.1 is certainly a plausible choice, because once 1-E-0 is initiated, the RNO of step 4 directs a transition to ES-0.1, without any notes or cautions in the EOP about this particular case, where a transition to ES-0.1 is NOT desired.
B. INCORRECT. See analysis for A. above. Although AP-24.01 has specific steps to ensure the proper SJAE alignment, a note before step 1 of AP-24.01 specifically states that ES-0.1 must NOT be performed in parallel.
C. CORRECT. Even though 1-E-0 step 4 RNO directs a transition to 1-ES-0.1, the correct flow path is to transition from 1-E-0 to 1-AP-24.01. This is specified in AP-24.00, which has as step 2, "Initiate 1-E-0..." and as step 3, "GO TO 1-AP-24.01...." In 1-AP-24.01, step 13 RNO will realign the correct valves and ensure the automatic actions take place.
D. INCORRECT. Transitioning to 1-AP-24.01 is correct; however, one should not carry out AP-24.00 actions in parallel with AP-24.01. Step 3 of AP-24.00 specifies that if a Reactor trip is required, the operator must initiate 1-E-0 and GO TO 1-AP-24.01--that is, one is NOT to remain in AP-24.00. Once a reactor trip occurs and 1-AP-24.01 is entered, there is no other (re-)entry condition into AP-24.00.
NOTE: another possible wrong distractor could be "operators are required to be able to correct a radiation monitor automatic action failure from memory ("skill of the craft")" for the second part of choices "B" and "D;" see Lesson Plan ND-93.5-LP-1-DRR.
Supporting References AP-24.00, "MINOR SG TUBE LEAK," rev 10, p. 2 and 3. AP-24.01, "LARGE STEAM GENERATOR TUBE LEAK," rev 28, p. 2 and 7 E-0, "REACTOR TRIP OR SAFETY INJECTION," rev. 61, p. 3
-Surry lesson plan ND-93.5-LP-1, "PRE-TMI RADIATION MONITORING SYSTEM," rev 10, p. 2, 16, slide 7 References Provided to Applicant
none OK with modication of (1). I.E. State transition to ES-0.1 and the AP-24.01 for C/D.
Answer: C
- 4. 003AG2.4.31 9 Unit 1 Initial Conditions:
Reactor power = 100%
Control rod D-6 rod bottom light lit 1G-H2, RPI ROD BOTTOM < 20 STEPS lit Control Rod D6 has been confirmed at the bottom of the core 0-AP-1.00 ROD CONTROL SYSTEM MALFUNCTION is entered Which ONE of the following correctly states (1) which procedure will be required to reduce reactor power and (2) the parameter that is required to be monitored to reduce and stabilize power?
A. (1) 1-GOP-2.1 (Unit Shutdown Power Decrease from Allowable Power to less that 30% Reactor Power)
(2) Loop T B. (1) 1-GOP-2.1 (Unit Shutdown Power Decrease from Allowable Power to less that 30% Reactor Power)
(2) the highest reading PRNI C. (1) 1-AP-23.00 (Rapid Load Reduction)
(2) Loop T D. (1) 1-AP-23.00 (Rapid Load Reduction)
(2) the highest reading PRNI K/A Dropped Control Rod: Knowledge of annunciator alarms, indications, or response procedures.
K/A Match Analysis Requires knowledge of response procedures for a dropped control rod.
SRO-Only Analysis Requires assessing plant conditions and then prescribing a procedure or section of a procedure to mitigate, recover, or with which to proceed. Knowledge above knowing entry conditions for APs is required.
Answer Choice Analysis A. Incorrect; 1st part is incorrect because AP/1.00 does not reference AP/23 and AP/1.00 gives an hour to reduce power to 70-74%. 1st part is plausible because AP/23 is frequently used to reduce power during plant upsets. 2nd part is correct per a caution in AP/1.00 before step 17.
B. Incorrect; 1st part is incorrect because AP/1.00 does not reference AP/23 and AP/1.00 gives and hour to reduce power to 70-74%. 1st part is plausible because AP/23 is frequently used to reduce power during plant upsets.2nd part is incorrect because caution in AP/1.00 states that DT must be monitored during the ramp and used to stabilize power. 2nd part is plausible because the highest reading PRNI will be more conservative than DT.
C. Correct: 1st part is AP/1.00 Step 17. A caution in AP/1.00 states that DT must be monitored during the ramp and used to stabilize power.
D. Incorrect; 1st part is correct. 2nd part is incorrect because caution in AP/1.00 states that DT must be monitored during the ramp and used to stabilize power. 2nd part is plausible because the highest reading PRNI will be more conservative than DT.
Supporting References 0-AP-1.00, ROD CONTROL SYSTEM MALFUNCTION References Provided to Applicant none Licensee discuss the potential use of AP/23 for the power reduction.
Good question, if 'A' was the correct answer. Even though there is not a step in AP-1.00 to direct entry into AP-23.00, that is our transient method to reduce power. Recommend changing stem slightly and selecting 'A' as the correct answer.
Answer: C
- 5. 0054G2.2.25 1 Which ONE of the following correctly identifies two reasons for the Feedwater Line Isolation function in response to a Safety Injection signal, as specified in the bases of Technical Specification 3.7, "INSTRUMENTATION SYSTEMS?"
A. (1) Prevent excessive cooldown of the Reactor Coolant System; AND (2) Reduces the consequences of a design basis steam generator tube rupture by preventing steam generator overfill.
B. (1) Prevent excessive moisture carry-over that could damage the main turbine blading; AND (2) Reduces the consequences of a design-basis steam generator tube rupture by preventing steam generator overfill.
C. (1) Prevent excessive cooldown of the Reactor Coolant System; AND (2) Reduces the consequences of a steam line break inside the containment by stopping the entry of main feedwater.
D. (1) Prevent excessive moisture carry-over that could damage the main turbine blading; AND (2) Reduces the consequences of a steam line break inside the containment by stopping the entry of main feedwater.
K/A Loss of Main Feedwater:
Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
(CFR: 41.5 / 41.7 / 43.2) (SRO - 4.2)
K/A Match Analysis The question is a straighforward link directly to the TS basis for feedwater isolation.
SRO-Only Analysis See attached SRO-only flowchart. TS Basis knowledge required to arrive at correct answer.
Answer Choice Analysis A. INCORRECT. The distractors are basically reasons for the HI-HI S/G level automatic function, worded to sound like the correct answers from the TS basis.
B. INCORRECT. see analysis of A. and C.
C. CORRECT. Answer is basically word-for-word from TS 3.7, which states:
"The feedwater lines are isolated upon actuation of the SIS in order to prevent excessive cooldown of the Reactor Coolant System. This mitigates the effects of an accident such as a steam line break which in itself causes excessive temperature cooldown. Feedwater line isolation also reduces the consequences of a steam line break inside the containment by stopping the entry of feedwater."
D. INCORRECT. See analysis of A. and C.
Supporting References
-Surry Technical Specification 3.7, amendment nos. 180 and 180, p. 3.7-5 and 3.7-6 References Provided to Applicant none
'A' is also correct for isolation due to HI HI level in accordance with the TS basis.
Answer: C
- 6. 0055G2.4.6 1 Unit 1 Initial Conditions:
A steam generator tube rupture caused an automatic reactor trip and SI from 100% power.
Operations personnel are performing actions in 1-E-3, "STEAM GENERATOR TUBE RUPTURE."
Current conditions:
A maximum-rate cooldown using steam dumps to the condenser has begun.
SI has just been reset.
The RO reports that condenser vacuum is 28 " Hg and slowly lowering.
The TSC informs the operations team that once all actions of E-3 are complete, it is required to implement the post-SGTR procedure that allows the FASTEST means of depressurizing the RCS and ruptured S/G.
Based on the current conditions, which ONE of the following is (1) a required action specified by E-3, AND (2) the correct post-SGTR procedure to implement?
A. (1) Ensure the condenser air ejector is aligned to containment, and then OPEN 1-SV-TV-102A.
(2) GO TO 1-ES-3.2, "POST-SGTR COOLDOWN USING BLOWDOWN."
B. (1) Ensure the condenser air ejector is aligned to containment, and then OPEN 1-SV-TV-102A.
(2) GO TO 1-ES-3.3, "POST-SGTR COOLDOWN USING STEAM DUMP."
C. (1) IF a Hi-CLS signal is NOT actuated, THEN realign the condenser air ejector for normal operations.
(2) GO TO 1-ES-3.2, "POST-SGTR COOLDOWN USING BLOWDOWN."
D. (1) IF a Hi-CLS signal is NOT actuated, THEN realign the condenser air ejector for normal operations.
(2) GO TO 1-ES-3.3, "POST-SGTR COOLDOWN USING STEAM DUMP."
K/A 055 Condenser Air Removal Knowledge of EOP mitigation strategies. (as relating to the Condenser Air Removal system)
(CFR: 41.10 / 43.5 / 45.13) (SRO - 4.7)
K/A Match Analysis The question requires the SRO applicant to demonstrate detailed knowledge of EOP mitigation strategies/transitions as related to expected effects of the condenser air removal system following an SI.
SRO-Only Analysis See attached SRO-only flowchart.
Linked to SRO-only knowledge based on detailed internal EOP transition criteria and procedural selection outside of initial/entry conditions.
Answer Choice Analysis A. INCORRECT. The lowering condenser vacuum is an expected condition. In the next few steps, 1-E-3 will ensure the proper operation of the air ejectors and mitigate the concern. Therefore the (1) part of this answer is correct. Part (2) is incorrect; the lesson plan for ES-3.3, "POST SGTR COOLDOWN USING STEAM DUMP," is very clear that it provides the fastest means of depressurizing the RCS and ruptured SG. ES-3.2 is plausible, if the applicant believes that the lowering condenser vacuum precludes the use of ES-3.3 through the steam dumps.
B. CORRECT. (1) Step 14 of 1-E-3 will align condenser air ejector to containment and improve the degraded vacuum condition. (2) is also correct; see analysis of A. above.
C. INCORRECT. (1) is incorrect, but plausible, because valve TV-SV-102 will (only) close automatically on a Hi-CLS signal. Also plausible because the question stem states that vacuum is lowering. Part (2) is also the incorrect procedural transition.
D. INCORRECT. (1) is incorrect choice, (2) is the correct proceural transition;
see above analyses.
Supporting References
-Surry lesson plan ND-89.3-LP-2, "MAIN CONDENSATE SYSTEM," rev. 18, p.
- 11. E-3, "STEAM GENERATOR TUBE RUPTURE," rev. 38, p. 10, 12.
-Surry lesson plan ND-95.3-LP-16, "ES-3.3 POST SGTR COOLDOWN USING STEAM DUMP," rev. 12, p. 31.
References Provided to Applicant none OK Answer: B
- 7. 006A2.12 12 Initial plant conditions on Unit 1 are as follows:
A SBLOCA has occurred.
Radiation levels in the Auxiliary Building are increasing.
The crew has transitioned to ECA-1.2 LOCA Outside Containment.
The crew closed/verified closed 1-SI-MOV-1890A and -1890B.
RCS pressure was at 1700 psig and slowly dropping.
Current plant conditions on Unit 1 are as follows:
The crew has closed 1-SI-MOV-1890C.
RCS pressure is at 1550 psig and slowly rising.
Which ONE of the following describes (1) the status of the LOCA and (2) the required procedure transition?
A. (1) LOCA has been isolated.
(2) Go to ECA-1.1, Loss of Emergency Coolant Recirculation.
B. (1) LOCA still exists.
(2) Go to ECA-1.1, Loss of Emergency Coolant Recirculation.
C. (1) LOCA has been isolated.
(2) Go to 1-E-1, Loss of Reactor or Secondary Coolant.
D. (1) LOCA still exists.
(2) Go to 1-E-1, Loss of Reactor or Secondary Coolant.
K/A Emergency Core Cooling: Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Conditions requiring actuation of ECCS.
K/A Match Analysis Requires applicant to predict the impact of a leak outside containment on the alignment of emergency core cooling equipment and perform the actions from ECA-1.2 for transitioning back to E-1.
SRO-Only Analysis The question requires the applicant to assess plant conditions and know the intent of the specific steps to determine the correct procedural transition..
Answer Choice Analysis A. In-Correct but plausible since the increasing RCS pressure indicates the leak has been isolated. In addition, the previous actions have closed all the cold and hot leg recirculation valves so it would seem plausible to transition to ECA-1.1, Loss of Emergency Coolant Recirculation. However, the correct action is to transition back to E-1.
B. In-Correct but plausible since the actions are correct if the leak still exists.
However, the increasing RCS pressure indicates the leak has been isolated and the crew should transition to E-1.
C. Correct - The increasing RCS pressure indicates the leak has been isolated.
The correct actions are to place LHSI pumps in PTL, close LHSI pump suction valves and transition to E-1.
D. In-Correct but plausible since reopening SI-MOV-1890C is correct if the leak still exists. The transition to E-1 is correct. However, the leak has been isolated.
Supporting References ND-95.3-LP-21, ECA-1.2, LOCA Outside Containment, Rev. 7, Obj. A References Provided to Applicant none NOTE:Original question used on Surry 02-301 exam - developed by G. Laska (WE04G2.4.9). Modified conditions to indicate isolation of leak and asked for
status of leak.
OK Answer: C
- 8. 0073A2.02 3 An electrical transient caused the failure of 1-RI-RM-159 (Containment Particulate Radiation Monitor) and 1-RI-RM-160 (Containment Gas Radiation Monitor).
Associated automatic actions have been verified.
Which ONE of the following states additional requirements of Annunciators 1-RM-Q7 (Containment Particulate Alert/Failure) and 1-RM-Q8 (Containment Gas Alert/Failure) for this failure?
A. A Technical Specification 3.0.1 LCO clock exists to return at least one of the radiation monitors to operable status.
B. Samples are required to be taken every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to ensure compliance with Technical Specification section 3.1.C (RCS Leakage).
C. As long as the manipulator crane, and incore area radiation monitors are operable, no other actions are required.
D. Within 7 days, establish an alternate method of detecting radionuclides in containment with remote indication in the main control room.
K/A Process Radiation Monitor (PRM) System Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
Detector failure.
(CFR: 41.5/43.5/45.3/45.13) (SRO - 3.2)
K/A Match Analysis Given a PRM detector failure condition, the SRO applicant will correctly determine the tech spec actions required.
SRO-Only Analysis This is an analysis level question since the candidate predict the consequence of the failure of the detector and impact on TS.
A- INCORRECT- plausible as tech specs 3.1.c requires that a method of radionuclide detection exist for RCS leak detection. Since both of those monitors are inoperable, candidate may believe that a 3.0.1. clock exists.
B- CORRECT C- INCORRECT - plausible as the candidate may believe that the other containment radiation monitors may suffice for the detection of conditions in containment.
D- INCORRECT - plausible as the TRM has many requirements to establish an alternate method of accomplishing compliance. Candidate may believe this to be a method of complying with detection requirements.
Supporting References
-modified from McGuire 2009-301 exam question SRO #94.
-Surry procedure 1A-A3, "N-16 HIGH," rev. 3.
-Surry procedure 1A-B3, "N-16 ALERT," rev. 3.
-Surry procedure 1A-C3, "N-16 TROUBLE," rev. 3.
References Provided to Applicant none Need to fix, our N-16 values do not change.
Answer: B
- 9. 0076AA2.02 1 Unit 1 Initial Conditions:
At time 0930, unexpected grid fluctuations caused an automatic turbine trip from 100% power.
Chemistry personnel drew a post-trip RCS sample at time 1005.
Control room operators have stabilized the unit at 547 °F and normal operating pressure.
Current conditions:
At time 1045, a Chemistry supervisor reports that the post-trip RCS sample total specific activity reading is greater than the 100/(E bar) limit by 28%.
Based on the current conditions, which ONE of the following (1) is the correct time the LCO for Technical Specification (TS) 3.1.D, Maximum Reactor Coolant Activity, is NOT met; AND (2) the basis of the requirement to cool down the reactor to less than 500 °F,
in accordance with TS 3.1.D?
A. (1) LCO not met at 1005; (2) In the unlikely event of an assumed 30 minute radioactive release during the design-basis S/G tube rupture, the iodine partitioning factor below this RCS temperature ensures exposure limits are not exceeded at the site boundary.
B. (1) LCO not met at 1045; (2) In the unlikely event of a design-basis S/G tube rupture, the saturation pressure corresponding to this RCS temperature is well below the pressure at which the atmospheric relief valves on the secondary side would be actuated.
C. (1) LCO not met at 1045; (2) In the unlikely event of an assumed 30 minute radioactive release during the design-basis S/G tube rupture, the iodine partitioning factor below this RCS temperature ensures exposure limits are not exceeded at the site boundary.
D. (1) LCO not met at 1005; (2) In the unlikely event of a design-basis S/G tube rupture, the saturation pressure corresponding to this RCS temperature is well below the pressure at which the atmospheric relief valves on the secondary side would be actuated.
K/A High Reactor Coolant Activity Ability to determine and interpret the following as they apply to High Reactor Coolant Activity: Corrective actions required for high fission product activity in RCS.
(CFR: 43.5/45.13) (SRO - 3.4)
K/A Match Analysis The question requires the SRO applicant to correctly demonstrate knowledge of the Technical Specifications for RCS activity, as well as the basis for this specification.
SRO-Only Analysis See attached SRO-only flow chart. TS Basis knowledge needed to arrive at correct answer.
Answer Choice Analysis A. INCORRECT. 1005 is the incorrect time, because the initial notification of the abnormality is considered the "start time" of inoperability. The second part of the answer is also incorrect; TS 3.1.D. basis states "Rupture of a steam generator tube would allow radionuclides in the reactor coolant to enter the secondary system. The limiting case involves a double-ended tube rupture coincident with loss of the condenser and release of steam from the secondary side to the
atmosphere via the main steam safety valves or atmospheric relief valves. This is assumed to continue for 30 minutes in the analysis. The operator will take action to reduce the primary side temperature to a value below that corresponding to the relief or safety valve setpoint. Once this is accomplished the valves can be closed and the release terminated." However, the distractor is plausible, because everything associated with this specification is concerned with a release during a design basis tube rupture.
B. CORRECT. See above analysis. The statement about the saturation pressure and atmospheric relief valves is basically word-for-word from the TS.
C. INCORRECT. Incorrect time, wrong reason for RCS cooldown.
D. INCORRECT. See above analysis.
Supporting References
-SPS TS 3.1.D References Provided to Applicant Steam Tables OK Answer: B
- 10. 010G2.4.20 13 Unit 1 initial conditions:
Reactor power = 100%
SGTR = 275 gpm on 1A SG Reactor is manually tripped 1C RCP trips Current conditions:
1-E-3 (STEAM GENERATOR TUBE RUPTURE) is in progress It is determined that Pzr spray is not adequately reducing RCS pressure and the decision is made to use the Pressurizer PORV to reduce RCS pressure.
Based on the above conditions, which ONE of the following states: (1) the reason for minimizing the cycling of the Pressurizer PORV and (2) the procedure that 1-E-3 directs you to perform if, during the depressurization, the Pressurizer PORV and its associated block valve CAN NOT be CLOSED?
A. (1) To reduce the chance of Pressurizer PORV failure.
(2) 1-ECA 3.3 SGTR WITHOUT PRESSURIZER PRESSURE CONTROL.
B. (1) To reduce the chance of Pressurizer PORV failure.
(2) 1-ECA 3.1 SGTR WITH LOSS OF REACTOR COOLANT - SUBCOOLED RECOVERY.
C. (1) To prevent the Tube rupture from degrading.
(2) 1-ECA 3.3 SGTR WITHOUT PRESSURIZER PRESSURE CONTROL.
D. (1) To prevent the Tube rupture from degrading .
(2) 1-ECA 3.1 SGTR WITH LOSS OF REACTOR COOLANT - SUBCOOLED RECOVERY.
K/A Pressurizer Pressure Control: Knowledge of the operational implications of EOP warnings, cautions, and notes.
K/A Match Analysis Requires knowledge of EOP Cautions.
SRO-Only Analysis Requires detailed knowledge of EOP steps having to do with securing PORV use when depressurizing the RCS.
Answer Choice Analysis A Incorrect: 1st part is correct. 2nd part is plausible because it is criteria for closing the PORV if Pzr level is > 22%.
B Correct: The PORV relieves to the PRT so using the PORV will eventually cause the PRT rupture disk to rupture. Criteria for securing from using the PORV are:
Pzr level>69%
RCS subcooling < 30 0F RCS press < Ruptured SG press AND Pzr level > 22%
C Incorrect: 1st part is plausible because the PORVs have failed to reseat (TMI) which constitutes a SBLOCA. 2nd part is plausible because it is criteria for closing the PORV if Pzr level is > 22%.
D Incorrect: 1st part is plausible because the PORVs have failed to reseat (TMI) which constitutes a SBLOCA. 2nd part is correct.
Supporting References 1-E-3 Steam Generator Tube Rupture. ND-95.3-LP-13 Obj A & B
References Provided to Applicant none Change a/b (1) to "to reduce chance of valve failure..."
Answer: B
- 11. 015/17AG2.2.22 2 Initial plant conditions on Unit 2 are as follows:
A power increase is in progress following reactor startup.
Reactor power is at 8%.
Pressurizer Spray valve 2-RC-PCV-2455A cannot be opened.
All three RCPs are operating.
Normal Charging has been tagged out at 1-CH-MOV-1289A and excess letdown is in-service due to required charging line piping repairs.
Current plant conditions on Unit 2 are as follows:
RCP C trips on ground overcurrent.
Based on the above conditions, which ONE of the following describes whether action statements of the following LCOs are required to be performed:
Technical Specification Section 3.1.A.4 (Reactor Coolant Loops)
Technical Specification Section 3.1.A.5 (Pressurizer)
Action statements within A. Technical Specification Section 3.1.A.4 is required.
Technical Specification Section 3.1.A.5 is NOT required.
B. Technical Specification Section 3.1.A.4 is NOT required.
Technical Specification Section 3.1.A.5 is required.
C. both Technical Specification Section 3.1.A.4 and 3.1.A.5 are required.
D. neither Technical Specification Section 3.1.A.4 nor 3.1.A.5 are required.
K/A RCP Malfunctions Knowledge of limiting conditions for operations and safety limits as it relates RCP Malfunctions.
K/A Match Analysis Applicant must recognize that loss of RCP C will require entry into both LCO 3.1.A.4. and 3.1.A.5.
SRO-Only Analysis The question requires a knowledge of the T.S. bases associated with LCO 3.1.A.4 concerning what constitutes an in-service reactor coolant loop to determine whether actions from LCO 3.1.A.4 are required.
Answer Choice Analysis A. In-Correct but plausible since LCO 3.1.A.4 would be entered given that LCO 3.1.A.4.b. states, POWER OPERATION with less than three loops in service is prohibited.. However, LCO 3.1.A.5 would also be entered since LCO 3.1.A.5.a states, The reactor shall be maintained subcritical by at least 1% until the steam bubble is established and the necessary sprays and at least 125 KW of heaters are operable. With PCV-455A inoperable, PCV-455B becomes inoperable once RCP C trips.
B. In-Correct but plausible if the applicant believes that a running RCP is not required for an RCS loop to be considered in service. The second half of the answer is correct. LCO 3.1.A.5 would be entered since LCO 3.1.A.5.a states, The reactor shall be maintained subcritical by at least 1% until the steam bubble is established and the necessary sprays and at least 125 KW of heaters are operable.
C. Correct -. Both LCO 3.1.A.4 and LCO 3.1.A.5 would be entered. See previous distractor discussions for justification.
D. In-Correct but plausible if the applicant believes that a running RCP is not required for an RCS loop to be considered in service AND does not recognized that both Pressurizer Spray valves are inoperable once RCP C trips.
NOTE TO LICENSEE: The correct answer was based on discussions with facility SME. The Technical Specifications bases do not provide a specific discussion with regards to what constitutes a loop being in service per LCO 3.1.A.4. Please provide documentation as to what constitutes a loop being in service.
Also, neither LCO 3.1.A.5 nor its basis states that sprays have an impact on
Technical Specifications once the reactor is above 1% subcritical. Please provide documentation for pressurizer operability when sprays are unavailable once a steam bubble is established and power is above 1%
subcritical.
Supporting References Technical Specification 3.1.A Technical Specification 3.0.1 ND-88.1-LP-9, Technical Specifications Overview, Rev. 16, Obj. G References Provided to Applicant none Answer: C
- 12. 025AA2.05 4 Unit 1 initial conditions:
Time = 0800 Plant was on RHR following shutdown for refueling Containment Closure has been established SGs are not available RCS temperature = 190°F stable RHR flow = 2200 gpm RCS level = 12.5 feet and decreasing RVLIS Full Range = 47% and decreasing Current plant conditions:
Time = 0825 1-AP-27.00 (LOSS OF DECAY HEAT REMOVAL CAPABILITY) has been initiated RHR pumps have been secured due to vortexing RCS temperature = 205°F increasing Based on the above conditions: (1) Classify the event using the Emergency Plan and (2) once RHR level and flow has been restored, state the MAXIMUM cooldown rate allowed per 1-AP-27.00?
(Reference Provided)
A. (1) Alert (2) 25°F/Hr B. (1) Alert (2) 50°F/Hr
C. (1) Site Area Emergency (2) 50°F/Hr D. (1) Site Area Emergency (2) 25°F/Hr K/A Loss of RHR: Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Limitations on LPI flow and temperature rates of change.
K/A Match Analysis Requires knowledge of limits on cooldown rate during loss of decay heat removal and recovery.
Requires the ability to determine the emergency classification based on the reduction and eventual loss of RHR flow due to invetory loss and requires knowledge of plant cooldown limits once RHR is restored.
SRO-Only Analysis Requires in depth knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures.
Answer Choice Analysis A. Incorrect: 1st part is incorrect because CS2 (Loss of Reactor Vessel inventory affecting core decay heat removal capability) existed = SAE. 1st part is plausible because CA 2 (Loss of RCS inventory) and CA3 (Inability to maintain plant in cold shutdown with irradiated fuel in the Reactor Vessel) apply. 2nd part is incorrect because 50 0F/Hr is the rate used for recovery once RHR is re-established. It is plausible because 25 0F/Hr is the cooldown rate for natural circulation cooldown in Attachment 4 of AP/27.
B. Incorrect: 1st part is incorrect because CS2 (Loss of Reactor Vessel inventory affecting core decay heat removal capability) existed = SAE. 1st part is plausible because CA 2 (Loss of RCS inventory) and CA3 (Inability to maintain plant in cold shutdown with irradiated fuel in the Reactor Vessel) apply. 2nd part is correct per
1AP/27 , Step 27.
C. Correct: 1st part is incorrect because CS2 (Loss of Reactor Vessel inventory affecting core decay heat removal capability) existed = SAE. 2nd part is correct per 1AP/27 , Step 27.
D. Incorrect: 1st part is incorrect because CS2 (Loss of Reactor Vessel inventory affecting core decay heat removal capability) existed = SAE. 2nd part is incorrect because 50 0F/Hr is the rate used for recovery once RHR is re-established. It is plausible because 25 0F/Hr is the cooldown rate for natural circulation cooldown in Attachment 4 of AP/27.
Supporting References Surry Emergency Plan AP/27 (LOSS OF DECAY HEAT REMOVAL CAPABILITY)
References Provided to Applicant Emergency Plan Answer: C
- 13. 027AA2.15 5 Unit 1 initial conditions:
Time = 1000 Reactor power = 100%
1-RC-PORV-1455C (Pressurizer Pressure PORV) indicates open Both Pzr Spray valves indicate open RCS Pressure = 2200 psig decreasing 1-AP-31.00 (Increasing or Decreasing RCS Pressure) initiated Current conditions:
Time = 1001 Reactor Power = 97%
RCS Pressure = 2100 psig increasing Spray valve in MANUAL and closed 1-RC-PORV-1455C in MANUAL and closed Based on the above conditions, which ONE of the following correctly states: (1) the component that failed high and (2) the status of 1-RC-PORV-1455C operability per Technical Specifications?
A. (1) 1-RC-PT-1444 (Pressurizer Pressure Control)
(2) PORV is considered OPERABLE
B. (1) 1-RC-PT-1444 (Pressurizer Pressure Control)
(2) PORV is INOPERABLE C. (1) 1-RC-PT-1445 (Pressurizer Pressure Control)
(2) PORV is considered OPERABLE D. (1) 1-RC-PT-1445 (Pressurizer Pressure Control)
(2) PORV is INOPERABLE K/A Pressurizer Pressure Control System Malfunction . Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: Actions to be taken if PZR pressure instrument fails high K/A Match Analysis Requires knowledge of how instrument failure affects the Pzr pressure control system and actions to mitigate the event.
SRO-Only Analysis Requires ability to interpret plant conditions and select appropriate AP/EOP to mitigate the event.
Answer Choice Analysis A. Incorrect. 1st part is correct. 2nd part is incorrect because the PORV is not able to perform its Normal Function at power (prevent challenging the code safetys).
2nd part is plausible because it is still operable in MANUAL.
B. Correct. Indications are indicative of transmitter P-444 failed high. TS directs the Block Valve for that PORV to be closed which renders the PORV inoperable.
If the PORV was still operable, this action would not be required. In the TS Bases 3.1.5c, it states this action is taken when the PORV is Inoperable.
C. Incorrect 1st part is incorrect because this transmitter does not control all of the functions to create the parameters listed. It is plausible because P-445 controls a PORV and will cause RCS pressure to decrease. 2nd part is incorrect because the PORV is not able to perform its Normal Function at power (prevent challenging the
code safetys). 2nd part is plausible because it is still operable in MANUAL.
D. Incorrect: 1st part is incorrect because this transmitter does not control all of the functions to create the parameters listed. It is plausible because P-445 controls a PORV and will cause RCS pressure to decrease. 2nd part is incorrect because the PORV is not able to perform its Normal Function at power (prevent challenging the code safetys). 2nd part is correct.
Supporting References TS Section 3.1.4a ND-93.3-LP5, Pzr Press Control pg 11 Obj: C References Provided to Applicant none Licensee to determine operability of PORV Answer: B
- 14. 035A2.01 19 Given the following Plant Conditions:
RCS Pressure = 1500 psig and slowly decreasing RCS Tave = 427 °F and slowly decreasing Pressurizer Level = 0%
Steam Generator Parameters are as follows:
Pressure Level (Wide Range)
A SG 120 psig - decreasing 23% - decreasing B SG 650 psig - slowly decreasing 56%- slowly increasing C SG 650 psig - slowly decreasing 56%- slowly increasing The Air Ejector Radiation Monitor is in High Alarm Containment Pressure is 16 psia and slowly increasing Containment Particulate and Gas Radiation Monitors (1-RI-RM-159/160) are in High Alarm Based on the conditions above, which ONE of the following describes the expected procedural sequence for this event?
A. 1-E-3 (Steam Generator Tube Rupture) 1-E-2 (Faulted Steam Generator Isolation) 1-ECA-3.1 (SGTR with Loss of Reactor Coolant - Subcooled Recovery)
B. 1-E-2 (Faulted Steam Generator Isolation) 1-E-3 (Steam Generator Tube Rupture) 1-ECA-3.1 (SGTR with Loss of Reactor Coolant - Subcooled Recovery)
C. 1-E-3 (Steam Generator Tube Rupture) 1-E-2 (Faulted Steam Generator Isolation) 1-E-1 (Loss of Reactor or Secondary Coolant)
D. 1-E-2 (Faulted Steam Generator Isolation) 1-E-3 (Steam Generator Tube Rupture) 1-E-1 (Loss of Reactor or Secondary Coolant)
K/A Steam Generator: Ability to (a) predict the impacts of the following malfunctions or operations on the S/GS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Faulted or Ruptured S/Gs.
K/A Match Analysis Requires knowledge of procedures used to mitigate a Faulted SG.
SRO-Only Analysis Requires the candidate to assess plant conditions and determine the appropriate procedural sequence to combat the event in progress.
Answer Choice Analysis A. INCORRECT. Procedural endpoint is correct and procedures used are correct. E-3 and E-2 are reversed. As most actions contained in E-3 would accomplish the actions contained in E-2, the candidate may believe this flow path is acceptable.
B. CORRECT.
C. INCORRECT. E-2/E-3 procedure order is out of squence (refer to distractor
'A'). E-1 is plausible as E-1 is the normal exit path from E-2 D. Incorrect: E-2/E-3 is in correct order. E-1 is plausible since the leak in containment and the event is akin to a SBLOCA.
Supporting References 1-E-2 1-E-3 1-ECA-3.1 1-E-1 References Provided to Applicant None
Answer: B
- 15. 051G2.4.11 12 With the unit initially at 100% the following conditions are encountered:
Annunciator 1A-G1 (TRAVELING SCREENS HI DIFF LVL) was received The outside operator reports high delta-P on A and C high level screens Vacuum is 27.6 and degrading PCS indications show CW outlet temperatures are 95°F and increasing Main generator output is 843 MWe and decreasing The team has initiated 1-AP-14.00 (Loss of Main Condenser Vacuum).
Which ONE of the following states (1) the cause of the vacuum trend and (2) the procedure directed by 1-AP-14.00 to reduce turbine load?
A. (1) Loss of heat sink to the condenser.
(2) 1-OP-TM-005, Unit Ramping Operations B. (1) Air in-leakage into the condenser.
(2) 1-AP-23.00, Rapid Load Reduction.
C. (1) Loss of heat sink to the condenser.
(2) 1-AP-23.00, Rapid Load Reduction D. (1) Air in-leakage into the condenser.
(2) 1-OP-TM-005, Unit Ramping Operations K/A Loss of Condenser Vacuum: Knowledge of abnormal condition procedures.
K/A Match Analysis Requires mitigation strategy of procedure for loss of vacuum.
SRO-Only Analysis Requires ability to assess plant conditins and then prescribing a procedure or section of a procedure to mitigate, recover or with which to proceed.
Answer Choice Analysis A- INCORRECT - Part 1 CORRECT - Part 2 plausible as candidate may believe that 1-OP-TM-005 is appropriate procedure for load reduction.
B - INCORRECT - Part 1 INCORRECT but plausible as candidate may believe indications point to an air inleakage event. Part 2 is CORRECT.
C - CORRECT D - INCORRECT- Part 1 INCORRECT (see (b) above) Part 2 incorrect (see (a) above).
Supporting References 1-AP-14.00 ND-95.1-LP-6 Obj: B References Provided to Applicant None Licensee to determine how much of SEP to be provided.
Answer: C
- 16. 059G2.4.14 15 Unit 1 initial conditions:
A reactor trip from 100% power occurs due to a Loss of Offsite Power Both EDGs start but both output breakers trip on ground over-current TD AFW pump fails to start All three Steam Generators indicate 54% Wide Range Level The STA reports that a RED path on Heat Sink exists and that all other status trees are green.
Based on the above conditions, which ONE of the following correctly states (1) the procedure that is required to be entered and (2) the classification of this event?
A. (1) 1-ECA-0.0 (Loss of All AC Power)
(2) Alert B. (1) 1-FR-H.1 (Response to Loss of Secondary Heat Sink)
(2) Alert C. (1) 1-ECA-0.0 (Loss of All AC Power)
(2) Site Area Emergency D. (1) 1-FR-H.1 (Response to Loss of Secondary Heat Sink)
(2) Site Area Emergency K/A
Main Feedwater: Knowledge of general guidelines for EOP usage.
K/A Match Analysis Requires knowledge of how the EOP directs feedwater restoration after a loss of all feedwater.
SRO-Only Analysis Requires knowledge of the basis of the status trees and the classification requirements for a loss of heat sink condition.
Answer Choice Analysis A. INCORRECT: ECA-0.0 is the only procedure that can be implemented at this time. The status trees assume one emergency bus is powered; these conditions are not met and therefore FR-H.1 is not applicable. Part two is plausible as even though you are not in FR-H.1, the conditions are met and and the candidate may not believe two potential losses results in a site area emergency (or they may only identify the RCS as the barrier in jeopardy).
B. INCORRECT: Part 1 is plausible because conditions of FR-H.1 are met and the STA has reported the conditions are met. This is the only status tree and it is RED. Red status trees must be implemented upon transition from E-0 (which this would have been, refer to step 3 of E-0). Part two is plausible as even though you are not in FR-H.1, the conditions are met and and the candidate may not believe two potential losses results in a site area emergency (or they may only identify the RCS as the barrier in jeopardy).
C. CORRECT.
D. INCORRECT: Part 1 is plausible because conditions of FR-H.1 are met and the STA has reported the conditions are met. This is the only status tree and it is RED. Red status trees must be implemented upon transition from E-0 (which this would have been, refer to step 3 of E-0). Part 2 is correct.
Supporting References Ref:
1-FR-H.1 RESPONSE TO LOSS OF SECONDARY HEAT SINK 1-ECA-0.0 LOSS OF ALL AC POWER References Provided to Applicant none Answer: C
- 17. 062AA2.06 2
Initial plant conditions:
Unit 2 shutdown with fuel offloaded Unit 1 = 100% power Current plant conditions:
Annunciator 1D-G5 (SW OR CC PPS DISCH TO CHRG PPS LO PRESS) is in alarm 1-AP-12.00 (SERVICE WATER SYSTEM ABNORMAL CONDITIONS) has been initiated Unit 1 operating CHG pump bearing temperatures:
1420 = 170°F 1440 = 180°F 1450 = 195°F Based on the above conditions: (1) which ONE of the following states the EARLIEST time at which 1-AP-12.00 directs shifting the operating charging pump due to high bearing temperature, and (2) if all charging pumps are incapable of providing flow, correctly states the reason for utilizing a Unit 2 Charging Pump in accordance with Technical Specifications?
A. (1) 1440 (2) To bring the operating unit to cold shutdown B. (1) 1440 (2) To bring the operating unit to hot shutdown ONLY C. (1) 1450 (2) To bring the operating unit to cold shutdown D. (1) 1450 (2) To bring the operating unit to hot shutdown ONLY K/A Loss of Nuclear Svc Water:
The length of time after the loss of SWS flow to a component before that component may be damaged.
K/A Match Analysis Requires knowledge of temperature limits on components supplied by SWS.
SRO-Only Analysis Requires knowledge of Tech Spec bases that is required to analyze Tech Spec required actions and terminology.
Answer Choice Analysis
A. Correct: At 180°F, AP/12 directs the charging pumps to be shifted. Per TS 3.2 C&VCS for a shutdown unit, one charging pump with a source of borated water shall be available for cross-connect with the operating unit so that if the operating units charging pumps become inoperable, the shutdown units charging pump can bring the disabled unit to cold shutdown.
B. Incorrect: 1st part is correct because at 180°F, AP/12 directs the charging pumps to be shifted. 2nd part is not correct because TS 3.2 states the shutdown units charging pump is used to bring the diabled unit to cold shutdown. 2nd part is plausible because being in hot shutdown would put the plant in a stable condition while repairs are conducted.
C. Incorrect: 1st part is incorrect because per AP/12 directs them to be shifted at 180°F. 1st part is plausible because at 195°F is a familiar number for RCP bearing temperatures. Part 2 is correct.
D. Incorrect: 1st part is incorrect because per AP/12 directs them to be shifted at 180°F. 1st part is plausible because at 195°F is a familiar number for RCP bearing temperatures. 2nd part is not correct because TS 3.2 states the shutdown units charging pump is used to bring the diabled unit to cold shutdown. 2nd part is plausible because being in hot shutdown would put the plant in a stable condition while repairs are conducted.
Supporting References TS 3.2, AP/12 Step 4 & 5, ND-89.5-LP-2 Obj H References Provided to Applicant none K/A - Having difficulty understanding part 2.
Answer: A
- 18. 079G2.2.22 2 Given the following plant conditions:
Unit 1 is at 100%
A loss of Containment Instrument Air has occurred 1B-F6, CTMT INST AIR HDR LO PRESSURE, annunciates
1D-C6, PRZR PWR RELIEF VV LO AIR PRESS, annunciates Containment Instrument Air was crosstied with Instrument Air Containment Instrument Air Pressure = 85 psig and increasing All Pressurizer PORV air bottles are properly aligned with air pressures of 1050 psig Which ONE of the following correctly states (1) the status of LCO 3.1.A.6 (PORV Operability) and (2) the Technical Specification required operator actions, if any?
A. (1) The LCO is met.
(2) No further action associated with the Pressurizer PORVs is required.
B. (1) The LCO is met.
(2) Verify Pressurizer PORV operability by closing Pressurizer PORV Block Valves, manually cycle the Pressurizer PORVs, and then re-open the Pressurizer PORV Block Valves.
C. (1) The LCO is NOT met.
(2) Restore the Pressurizer PORV backup air supply within 14 days OR be in HSD within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. (1) The LCO is NOT met.
(2) Close and remove power from both Pressurizer PORV block valves within one hour AND be in HSD within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 079 Station Air G2.2.22: Knowledge of limiting conditions for operations and safety limits K/A MATCH ANALYSIS:
The question requires knowledge of PORV operability which is impacted by a loss of air. The operability determination causes the conditions of the LCO to not be met.
SRO-ONLY ANALYSIS:
Operability is primarily an SRO function unless the determinatation is made at a very basic level (I.E. if a pump is broke, it is obviously inop - which would be RO knowledge). This question requires the SRO to understand how the loss of instrument air affects the PORV operability, even when the PORV is available for use with cross-tied air.
Answer Choice Analysis:
A. Incorrect per 1D-C6 CTMT Inst Air P must be > 80 psig for the PORVs to be operable.
B. Incorrect because (per 1D-C6) with CTMT Inst Air P < 80 psig, the PORVs are inoperable.
C. Correct because PORVs are capable of being manually cycled with CTMT Inst Air P > 80 psig. The PORVs are INOP due to INOP air supply and you start a 14 day LCO clock.
D. Incorrect, the PORV is INOP but can be manually cycled. This choice is correct if the PORV could NOT be manually cycled. This would be a 1 hr LCO.
Surry Requal Bank Question #571 (LARP0001) & 2004-301 NRC Exam
References:
ND-92.1-LP-1, Station Air Systems, Rev. 13 ND-88.1-LP-3, Pressurizer and Pressure Relief, Rev. 12 1B-F6, CTMT INST AIR HDR LO PRESS, Rev. 1 1D-C6, PRZR PWR RELIEF VV LO AIR PRESS, Rev. 4 Technical Specification 3.1.A.6.c, Reactor Coolant System / Relief Valves Containment IA is stated at 85 psig.
Question confuses pressures (80 and 85) and confuses TRM and TS clocks.
Question confuses containment IA with przr porv regulator pressure.
Answer: A
- 19. G2.1.20 14 Initial plant conditions on Unit 2 are as follows:
Reactor power is 100%.
A 20 gpd leak exists on steam generator B.
Current plant conditions on Unit 2 are as follows:
Charging flow has slowly increased.
Auto-makeup to VCT has started.
VCT level is 29% and slowly rising.
Pressurizer level is stable at 54%.
Pressurizer pressure is stable at 2225 psig.
Crew has entered 1-AP-16.00 (Excessive RCS Leakage).
Radiation levels on MSL B show a slow increasing trend.
The leak rate has been calculated at 12 gpm.
Transition to 2-AP-24.00, Minor SG Tube Leak is (1) AND the correct classification (if any) for the event is a (an) (2) ?
(Reference provided)
A. (1) required (2) None Required B. (1) NOT required (2) None Required C. (1) required (2) NOUE D. (1) NOT required (2) NOUE
[DISCUSS WITH THE LICENSEE TO DETERMINE CONDITIONS FOR THE STEM THAT WILL ENSURE ONE AND ONLY ONE CORRECT ANSWER AS WELL AS PLAUSIBILITY FOR THE DISTRACTORS]
K/A Generics: Ability to interpret and execute procedure steps.
K/A Match Analysis Requires applicant to interpret the leak indications, determine if transition to 1-AP-24.00 is required and determine the correct emergency classification associated with the leak.
SRO-Only Analysis The question requires the applicant to correctly determine if a procedure transition is required from AP-16-00 and classify the event per the emergency plan. Both of which would require SRO- Only knowledge to determine.
Answer Choice Analysis A. In-Correct but plausible since a procedure transition to 1-AP-24.00 is required.
B. In-Correct but plausible since a procedure transition would not be required if the applicant didn't recognize that MSL 'B' radition were increasing.
C. Correct - Transition to 1-AP-24.00 is required.
D. In-Correct. See above.
Supporting References 1-AP-16.00, Excessive RCS Leakage, Rev. 16 Emergency Plan, Rev. 54 References Provided to Applicant
Emergency Plan NOTE: Facility reviewers please validate that the correct emergency classification was determined.
OK Answer: C
- 20. G2.2.14 22 Plant conditions:
RCS cooldown in progress RCS temperature = 350°F decreasing RCS pressure = 300 psig Based on the above conditions, with regards to the Over Pressure Mitigation System (OPMS), prior to decreasing below 350°F (1) , and the TS basis for that configuration is (2) ?
(Consider No TS modifications, LCOs...)
A. (1) Pzr level is limited to 33% for the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2) To allow the operator 10 minutes to take action from inadvertent initiation of full (3 pump) charging flow.
B. (1) Two PORVs are required to remain operable (2) Based on the Pressurizer PORVs ability to relieve RCS pressure from the start of a RCP with SG temp > RCS temp.
C. (1) Accumulators must be depressurized to less than the Pressurizer PORV setpoint (2) To prevent exceeding the PORV capability if an inadvertent OPMS initiation occurs.
D. (1) Verify a maximum of one charging pump is capable of injecting into the RCS.
(2) To ensure any mass addition can be relieved by one PORV.
K/A Knowledge of the process for controlling equipment configuration or status.
K/A Match Analysis Requires knowledge of the equipment configuration for specific plant conditions.
SRO-Only Analysis Requires knowledge of the plant configuration for cooldown operations and the TS Bases for that configuration.
Answer Choice Analysis A. Incorrect: Plausible because the limit is correct but based on only one charging pump injecting.
B. Incorrect: 2 PORVs are required for the 1st 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if no vent exists or Pzr level
< 33%. Plausible because the bases stated is for one PORV being operable.
C. Incorrect: Accumulators can be isolated and valves de-energized as an alternative to depressurizing. While initiation may cause the PORV to lift, it will not exceed its capacity. Plausible because depressurizing the accumulators is an option to isolating them.
D. Correct. Per TS 3.1.G Supporting References ND-93.3-LP-6 Obj: E TS3.1.G References Provided to Applicant none OK Answer: D
- 21. G2.2.22 4 In accordance with Technical Specification section 3.16 (Emergency Power System),
which ONE of the following (1) states the MINIMUM level of fuel oil in the underground fuel oil storage tanks, and (2) the basis for this requirement?
A. (1) 20,000 gallons (2) To support the full load operation of one EDG for 7 days.
B. (1) 20,000 gallons (2) To support the full load operation of three EGDs for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
C. (1) 35,000 gallons (2) To support the full load operation of one EDG for 7 days.
D. (1) 35,000 gallons (2) To support the full load operation of three EGDs for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
K/A Knowledge of limiting conditions for operations and safety limits.
(CFR: 41.5/43.2/45.2) (SRO - 4.7)
K/A Match Analysis The K/A is a Tier 3, or "generic" K/A. The question asks the SRO candidate to demonstrate knowledge of the bases for an important Technical Specifications LCO for emergency power distribution.
SRO-Only Analysis
- "Below the LIne" TS knowledge is required to answer this question.
Answer Choice Analysis A. INCORRECT. Part one is incorrect, but plausible as it states the capacity of the tanks, rather than the required level. Part 2 is correct.
B. INCORRECT. See (A) for part (1) plausibility. Part 2 is plausible as 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is usually the time associated with obtaining CSD.
C. CORRECT.
D. INCORRECT. Part (1) correct. Part (2) plausible - see (B) above.
Supporting References SPS TS 3.16 and bases.
References Provided to Applicant None Answer: C
- 22. G2.3.12 1 Unit 1 initial conditions:
Date = 6/24 Time = 0800 Reactor power = 100%
Waste gas storage tank activity level is reported which exceeds TS 3.11, Radioactive Gas Storage, limits Current conditions:
Date = 6/26
Time = 0800 Reactor power = 100%
Waste gas storage tank activity level still exceeds Tech Spec 3.11 limits Based on the above conditions, which ONE of the following correctly states: (1) if Tech Spec 3.0.1 is applicable and (2) the whole body dose that the tank radioactivity limit is designed to prevent exceeding at the exclusion area boundary if the tank were released IAW Tech Spec Basis?
A. (1) Yes (2) 50 mrem B. (1) Yes (2) 0.5 rem C. (1) No (2) 50 mrem D. (1) No (2) 0.5 rem K/A Knowledge of radiological safety principles pertaining to licensed operator l duties, such as containment entry requirements, fuel handling responsibilities, l access to locked high-radiation areas, aligning filters, etc.
K/A Match Analysis Requires knowledge of radiological limits associated with the health and safety of the public and how to apply technical specifications to stay within those limits.
SRO-Only Analysis Requires knowledge of the facility operation limitations in the technical specifications and their bases.
Answer Choice Analysis A. Incorrect: In TS 3.11.B.3 it states that the requirements fo Specification 3.0.1 are not applicable. 1st part is plausible because the time for condition 3.11.B.2 has expired. 2nd part is incorrect because in the TS bases 3.11 it states 0.5 rem.
2nd part is plausible becasue the Surry adminestrative limit for site visitors is 50 mrem.
B. Incorrect: In TS 3.11.B.3 it states that the requirements fo Specification 3.0.1 are not applicable. 1st part is plausible because the time for condition 3.11.B.2
has expired. 2nd part is correct per TS 3.11 bases.
C. Incorrect: 1st part is correct. 2nd part is incorrect because in the TS bases 3.11 it states 0.5 rem. 2nd part is plausible becasue the Surry adminestrative limit for site visitors is 50 mrem.
D. Correct: In TS 3.11.B.3 it states that the requirements fo Specification 3.0.1 are not applicable. In the tech spec bases for TS 3.11 it states it limited to the quantity which provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem in an event.
Supporting References TS 3.11 ND-81.2-LP3 References Provided to Applicant none Needs more validation, something caused the question to be missed.
Answer: D
- 23. G2.3.4 22 Unit 1 initial plant conditions:
Reactor power = 50%
Plant shutdown in progress due to RCS activity greater than TS limits AFW Pump 1-FW-P-3B OOS Current plant conditions:
A SG tube rupture occurs
'A' SG pressure = 1000 psig Reactor has been tripped 1-E-3 STEAM GENERATOR TUBE RUPTURE in progress The TSC has been established An operator is dispatched to close 1-MS-87 (steam from the A SG to the TD AFW pump) in order to save valuable equipment Based on the above conditions, which ONE of the following: (1) states the MAXIMUM allowable dose (TEDE) that can be authorized for the operator to receive while isolating steam to the TD AFW pump and (2) if the valve can not be closed, what procedural actions shall be taken IAW 1-E-3 to mitigate the failure?
A. (1) 10 Rem (2) Remain in 1-E-3 and trip the TD AFW pump overspeed trip valve.
B. (1) 10 Rem (2) GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT -
SUBCOOLED RECOVERY.
C. (1) 25 Rem (2) Remain in 1-E-3 and trip the TD AFW pump overspeed trip valve.
D. (1) 25 Rem (2) GO TO 1-ECA-3.1, SGTR WITH LOSS OF REACTOR COOLANT -
SUBCOOLED RECOVERY.
K/A Knowledge of radiation exposure limits under normal or emergency conditions.
K/A Match Analysis Requires knowledge of exposure limits under emergency conditions.
SRO-Only Analysis Requires knowledge of EOP procedures and transition points.
Answer Choice Analysis A. Correct: Allowable dose for equipment = 10 Rem. Per 1-E-3, if at least 1 motor driven AFW pump available, trip the TD AFW pump.
B. Incorrect: 1st part is correct. 2nd part is plausible because if the SG with the rupture could not be isolated from both of the intact SGs, it would be correct.
C. Incorrect: 1st part is plausible because the exposure could be counted towards a PSE (the PSE limit is 5 Rem / yr). 2nd part is correct.
D. Incorrect: 1st part is plausible because the exposure could be counted towards a PSE (the PSE limit is 5 Rem / yr). 2nd part is plausible because if the SG with the rupture could not be isolated from both of the intact SGs, it would be correct.
Supporting References ND-81.2-LP-3 Obj: E 1-E-3 ND-95.3-LP-13 E-3 Obj: A
References Provided to Applicant none OK Answer: A
- 24. G2.4.30 3 Unit 1 Initial Conditions:
Holding at 50% power for fuel conditioning following a refueling outage.
Current conditions:
Technicians performing a routine surveillance test on the AMSAC logic system inadvertently cause a half-train Train "A" AMSAC signal to be generated.
Annunciator F-B-3, AMSAC INITIATED, is lit All AMSAC functions occurred.
Based on the current conditions, which ONE of the following correctly describes (1) whether the half-train AMSAC signal should be considered a VALID or INVALID actuation, as defined by VPAP-2802, "Notifications and Reports," AND (2) the most restrictive time requirement to report this event to the NRC, as specified by VPAP-2802?
(Reference provided)
A. (1) VALID actuation (2) 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification B. (1) INVALID actuation (2) 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification C. (1) VALID actuation (2) 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> notification D. (1) INVALID actuation (2) 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification K/A Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
(CFR: 41.10/43.5/45.11) (SRO - 4.1)
K/A Match Analysis The question requires the applicant to demonstrate knowledge of the definitions
inherent in the notifications procedure ("system operation/status"), and also show an ability to use the procedure to determine the correct time requirements for the given plant conditions, which are operationally valid.
SRO-Only Analysis This question requires the applicant to know the definitions inherent in the Notifications procedure, and to apply them in a practical setting. Therefore, it is a higher-level comprehension/analysis question that is linked to 10CFR55.43(b)(1),
"conditions and limitations in the facility license," in that ROs are not required to know and be able to apply reporting requirements.
Answer Choice Analysis A. INCORRECT. (1) Surry/Dominion procedure VPAP-2802, "Notifications and Reports," section 4.3 specifies that a VALID actuation must result "from an intentional manual initiation or from a signal that was initiated in response to actual plant conditions or parameters satisfying the requirements for initiation, unless part of a preplanned test." For the given conditions, the inadvertant AMSAC actuation was caused as a result of testing, the actuation was a result of human error and was not pre-planned to occur, and was not in response to actual plant conditions. Therefore, to state that the actuation was VALID is plausible. (2) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is the most restrictive notification, to report an RPS actuation on a critical reactor.
B. INCORRECT. (1) VPAP-2802 section 4.2 specifies that an invalid actuation "is one that does not meet the criteria for being valid and are initiated for reasons other than to mitigate the consequences of an event (e.g., as part of a planned evolution, with the system properly removed from service, or after the safety function has already been completed). Invalid actuations include circumstances where instrument drift, spurious signals, human error, or other invalid signals caused actuation (e.g. jarring a cabinet, an error in the use of jumpers or lifted leads, an error in the actuation of switches or controls, equipment failure, radio frequency interference)." For the given conditions, human error caused the actuation; therefore INVALID actuation is correct. The candidate must then infer from the question whether the reactor tripped (yes). (2) Based on the provided reference material, the candidate my incorrectly choose an 8-hour notification based on auxiliary feedwater auto-start, if he/she incorrectly believes that the AMSAC actuation at a low power level would not produce a reactor trip (or only trip the turbine and not the reactor as well). The plausibility of this choice is enhanced by the question stem stating that the signal is reset within 10 seconds (where a normal AMSAC signal is required to remain "in" for 27 seconds to cause an actuation).
C. INCORRECT. "VALID" actuation is wrong as per the above.
D. CORRECT. "INVALID" actuation is correct as per the above. VPAP-2802
section 6.3.4.a.3. states that a 4-hour report is required for "Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when actuation results from and is part of a pre-planned sequence during testing or reactor operation." In this case, an automatic reactor trip/RPS actuation did occur with the reactor critical. The reactor trip was not pre-planned; rather, it was caused by human error, and therefore the exclusion clause does not apply.
Supporting References
-VPAP-2802, "Notifications and Reports," rev 30, (p. 20, p. 82, and p. 86)
- Surry lesson plan ND-93.3-LP-17, "ANTICIPATORY MITIGATING SYSTEM ACTUATING CIRCUITRY (AMSAC)," rev. 11, p. 7 and 9.
References Provided to Applicant
-VPAP-2802, "Notifications and Reports," pages 79-91.
Need a reference.
Answer: D
- 25. G2.4.9 24 Unit 1 plant conditions:
Time = 0200 RCS cooldown in progress RCS temperature = 250 oF RCS pressure = 320 psig 1A charging pump is the only running charging pump Current plant conditions:
Time = 0210 RCS pressure = 280 psig decreasing The maximum charging flow achieved with the 1A charging pump is 125 gpm Based on the above conditions, which ONE of the following: (1) states the correct procedure to be entered and (2) what actions are directed by that procedure?
A. (1) 1-AP-16.00 EXCESSIVE RCS LEAKAGE (2) Align charging pump suction to the RWST B. (1) 1-AP-16.00 EXCESSIVE RCS LEAKAGE (2) Align and start 1B and 1C charging pumps
C. (1) 1-AP-16.01 SHUTDOWN LOCA (2) Align charging pump suction to the RWST D. (1) 1-AP-16.01 SHUTDOWN LOCA (2) Align and start 1B and 1C charging pumps K/A Knowledge of low power / shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.
K/A Match Analysis Requires knowledge of shutdown procedures/mitigation strategies during an accident.
SRO-Only Analysis Requires in depth knowledge of abnormal procedure guidelines and selection based on plant conditions.
Answer Choice Analysis A. Incorrect: Note at the top of AP/16.00 states If SI Accumulators are isolated, 1-AP-16.01, SHUTDOWN LOCA, should be used for guidance. Plausible because if > 350 0F, it would be correct. 2nd part is correct.
B. Incorrect: Note at the top of AP/16.00 states If SI Accumulators are isolated, 1-AP-16.01, SHUTDOWN LOCA, should be used for guidance. 2nd part is plausible because if > 350 0F charging pumps would used as necessary per AP/16.00 (OPMG not in service). Having OPMG in service requires only 1 Chg available to inject into the RCS.
C. Correct. If SI Accumulators are isolated, 1-AP-16.01, SHUTDOWN LOCA, should be used for guidance. Being < 350 oF requires the accumulators to be isolated.
2nd part is step 8 d RNO.
D. Incorrect: 1st part is correct. 2nd part is plausible because if > 350 0F charging pumps would used as necessary per AP/16.00 (OPMG not in service).
Having OPMG in service requires only 1 Chg available to inject into the RCS.
Supporting References Ref: AP/16.00, AP/16.01
References Provided to Applicant none OK Answer: C